International Conference
Nuclear Energy f0r New Eur0pe 2010
Book of Abstractst www.nss.si/port200
The contents of abstracts published in this book are the responsibility of the authors concerned. The organizer is not responsible for published facts and technical accuracy of presented data. Organizers would also like to apologize for any error caused by electronic transmission and materials processing.
_Previous meetings organized by Nuclear Society of Slovenia:
• First Meeting of Nuclear Society of Slovenia, Bovec, Slovenia, June 1992 • Regional Meeting: Nuclear Energy in Central Europe, Present and Perspectives, Portorož, Slovenia, June 1993 • PSA/PRA and Severe Accidents '94, Ljubljana, Slovenia, April 1994 • Annual Meeting of NSS '94, Rogaška Slatina, Slovenia, September 1994 • 2nd Regional Meeting: Nuclear Energy in Central Europe, Portorož, Slovenia, September 1995 • 3rd Regional Meeting: Nuclear Energy in Central Europe, Portorož, Slovenia, September 1996 • 4th Regional Meeting: Nuclear Energy in Central Europe, Bled, Slovenia, September 1997 • Nuclear Energy in Central Europe `98, Čatež, Slovenia, September 1998 • Nuclear Energy in Central Europe `99 with Embedded Meeting Neutron Imaging Methods to Detect Defects in Materials, Portorož, Slovenia, September 1999 • 20th International Conference on Nuclear Tracks in Solids, Portorož, Slovenia, August 2000 • Nuclear Energy in Central Europe 2000, Bled, Slovenia, September 2000 • Nuclear Energy in Central Europe 2001, Portorož, Slovenia, September 2001 • Nuclear Energy for New Europe 2002, Kranjska Gora, Slovenia, September 2002 • Nuclear Energy for New Europe 2003, Portorož, Slovenia, September 2003 • Nuclear Energy for New Europe 2004, Portorož, Slovenia, September 2004 • Nuclear Energy for New Europe 2005, Bled, Slovenia, September 2005 • Nuclear Energy for New Europe 2006, Portorož, Slovenia, September 2006 • Nuclear Energy for New Europe 2007, Portorož, Slovenia, September 2007 • Nuclear Energy for New Europe 2008, Portorož, Slovenia, September 2008 • Nuclear Energy for New Europe 2009, Bled, Slovenia, September 2009
_Sponsors
_Organizer Address
Contact Address: Nuclear Society of Slovenia, PORT2010 Jamova 39, SI-1000 Ljubljana Slovenia E-mail :[email protected] www: www.djs.si/port2010 Phone: +386 1 588 53 95 Fax: +386 1 588 53 77
_Program Committee
Andrej Trkov, Slovenia, Chair
Helmuth Böck, Austria Ivan Alexander Kodeli, Slovenia Leon Cizelj, Slovenia Alojz Kodre, Slovenia Gerard Cognet, France Boštjan Končar, Slovenia Nikola Čavlina, Croatia Božidar Krajnc, Slovenia Marko Čepin, Slovenia Martin Kropík, Czech Republic Milan Čerček, Slovenia Vladislav Krošelj, Slovenia Duško Čorak, Croatia Andrej Likar, Slovenia Oscar Cabellos de Francisco, Spain Borut Mavko, Slovenia Francesco D'Auria, Italy Irena Mele, IAEA Milorad Dušič, IAEA Martin Novšak, Slovenia Danilo Feretić, Croatia Primož Pelicon, Slovenia Paolo Gaio, Belgium Dubravko Pevec, Croatia Michel Giot, Belgium Stane Rožman, Slovenia Horst Glaeser, Germany Rainer Salomaa, Finland Bogdan Glumac, Slovenia Vladimír Slugeň, Slovak Republic Davor Grgić, Croatia Andrej Stritar, Slovenia Yassin Hassan, USA Iztok Tiselj, Slovenia Tim Haste, UK Zoran V. Stošić, Germany Igor Jenčič, Slovenia Pedro Vaz, Portugal Romana Jordan Cizelj, Slovenia Tomaž Žagar, Slovenia Ivo Kljenak, Slovenia
_Best Paper Award Committee
Alojz Kodre, Slovenia, Chair
Francesco D’Auria, Italy Michel Giot, Belgium
_Best Poster Award Committee
Tim Haste, France, Chair
Gerard Cognet, France Igor Jenčič, Slovenia
_Organizing Committee
Igor Lengar, Chair
Robert Bergant Dušan Rudman Slavko Slavič Darja Stich Urška Turšič Bojan Žefran Gašper Žerovnik
_Preliminary Conference Program
7.9.2010
Session 3 9:20 Invited Lecture 1
9:20 1 Robin Forrest - Austria The Role of Nuclear Data for Fusion Technology Studies
Session 9 10:00 Reactor Physics, Fuel Cycle and Research Reactors
10:00 204 P. Dossantos-Uzarralde, H.P. Jacquet, G. Gauriot - France Methodology Investigations on Uncertainties Propagation in Nuclear Data Evaluation 10:20 209 Marco Pecchia, G. Kotev, Carlo Parisi, Francesco D'Auria - Italy MOX Benchmark Calculation Results by Monte Carlo and Deterministic Codes 10:40 202 Nikolas Catsaros, B. Gaveau, M. Jaekel, Jacques Maillard, G. Maurel, P. Savva, J. Silva, M. Varvayanni - Greece Building a Dynamic Monte Carlo Code to Simulate New Reactor Concepts
Session 11 11:20 Thermal Hydraulics
11:20 408 Laurent Bricteux, Matthieu Duponcheel, Yann Bartosiewitz - Belgium Direct Numerical Simulation of Heat Transfer at Very Low Prandtl Number: Application to convection convection in Lead-Bismuth flows
11:40 402 Eckhard Krepper, Martin Schmidtke, Dirk Lucas, Matthias Beyer - Germany Steam Bubble Condensation in Polydispersed Flow –Experiments and CFD Simulations 12:00 412 Yuriy Parfenov, Vladimir Melikhov, Oleg Melikhov, Alexey Nerovnov - Russia Analysis of the Two-phase Flow Phenomena at PGV Test Facility
Session 4 14:00 Invited Lecture 2
14:00 2 Enrico Sartori - France Models and Their Benchmarking - International Experience in Nuclear Applications
Session 13 14:40 Materials and Structural Integrity
14:40 604 Oriol Garrido, Leon Cizelj, Igor Simonovski - Slovenia Modular 3-D Solid Finite Element Model for Fatigue Analyses of a PWR Reactor Coolant System
15:00 610 Christian Poeckl, Steffen Bergholz, Juergen Rudolph, Nikolaus Wirtz - Germany AREVA’s Fatigue Concept (AFC) – An Integrated And Multidisciplinary Approach To The Fatigue Assessment Of NPP Components
15:20 607 Artur Muehleisen, Matjaž Podjavoršek, Andreja Peršič, Djordje Vojnovič, Andrej Stritar - Slovenia Slovenian Regulatory Approach to Design Lifetime Extension
Session 14 16:00 Severe Accidents and Probabilistic Safety Assessment
Probabilistic Safety Assessment 16:00 702 Duško Kančev, Marko Čepin - Slovenia Sensitivity and Uncertainty Analysis for Age-Dependent Model of Test and Maintenance
16:20 804 Tim Haste, Frédéric Payot, Cristina Dominguez, Philippe March, B. Simondi-Teisseire, Martin Steinbrück, Roland Zeyen - France Study of Boron Behaviour in the Primary Circuit of Water Reactors under Severe Accident Conditions; a Comparison of Recent Integral and Separate-Effect Data
16:40 801 Oleg Solovjanov, R. Lutz, Nathalie Dessars, Thibaut Rensonnet - Belgium Implementation of Severe Accident Management Guidelines to Shutdown and Low- Power Operation Modes for VVER and PWR Plants
Session 15 17:10 Radioactive Waste and Decommissioning
17:10 905 Ana Stanič - United Kingdom EU Law on Radioactive Waste 17:30 902 Boštjan Duhovnik, Janja Špiler - Slovenia Preliminary Design of Vrbina LILW Repository 17:50 903 Marko Kostanjevec, Marija Fabjan, Simona Sučić - Slovenia Transport of Radioactive Waste from Small Producers 18:10 908 Nadja Železnik, Metka Kralj, Irena Mele, Primož Stropnik, Ivica Levanat, Vladimir Lokner, Andrea Rapić - Slovenia Revision 2 of the Program of NPP Krško Decommissioning and SF & LILW Disposal
8.9.2010
Session 5 9:00 Invited Lecture 3
9:00 5 Cheng Xu Cross-Cutting Tasks on Thermal-Hydraulics of Innovative Nuclear Systems
Session 12 9:40 Multiphyiscs
9:40 505 György Hegyi, András Keresztúri, Csaba Maŕaczy, Istvan Trosztel - Hungary Reactivity Initiated Accident Analysis of the HPLWR Three Pass Core with Ascending Gap Flow Using the KIKO3D-ATHLET Code
10:00 501 R. Miró, C. Pereira, T. Barrachina, G. Verdú, J.C. Martínez-Murillo - Spain Implementation of New Simulation Capabilities in RELAP5/PARCS v2.7 Coupled Codes 10:20 504 Davide Rozzia, Martina Adorni, Alessandro Del Nevo, Francesco D'Auria - Italy TRANSURANUS Verification Against CNEA’s PHWR MOX Experiments, from IFPE Database
Session 20 11:00 In memory prof. Matjaž Ravnik
Reactor Physics, Fuel Cycle and Research Reactors 11:20 203 Helmuth Böck, Rustam Khan - Austria Neutronics Analysis of the TRIGA Vienna Mixed Core
Session 17 11:50 Radiation and Environment Protection
11:50 1004 Matjaž Koželj - Slovenia A Brief History of Krško NPP Radiation Impact on Environment 12:10 1103 Marko Černe, Borut Smodiš - Slovenia Comparison of Measured Activity Concentrations in Plants From the Area of the Former Uranium Mine at Žirovski Vrh, Slovenia with Activity Concentrations Obtained by the ERICA Biota Assessment Tool
9.9.2010
Session 6 9:00 Invited Lecture 4
9:00 3 Brian Syme - United Kingdom Fusion Yield Measurements on JET and their Calibration
Session 16 9:40 Nuclear Fusion and Plasma Technology
9:40 1010 Jesus Izquierdo - Spain ITER, an Essential Step in Fusion Development 10:00 1014 Hua Sheng, Inge Uytdenhouwen, Vincent Massaut, Guido Van Oost, Jozef Vleugels - Belgium Mechanical Properties and Microstructural Characterizations of Potassium Doped Tungsten
10:20 1002 Luka Snoj, Brian Syme, Sergey Popovichev, Igor Lengar, Sean Conroy - Slovenia Calculations to Support JET Neutron Yield Calibration: Contributions to the External Neutron Monitor Responses
Session 1 10:40 Posters
Nuclear Renaissance and Prospects of Nuclear Energy 10:40 102 Claudia Guerrieri, Antonio Cammi, Carlo Fiorina, Lelio Luzzi - Italy A Multi-Physics Numerical Model for the MSR Core Dynamics 10:40 105 Carlo Fiorina, Antonio Cammi, Claudia Guerrieri, Lelio Luzzi, Benedetto Spinelli - Italy Development of a Simulation Tool for a Preliminary Analysis of the MSR Core 10:40 106 Laure Lizon-A-Lugrin, Alberto Teyssedou, Igor Pioro - Canada Appropriate Thermodynamic Cycles to be Used in Future Pressure-Channel Supercritical Water-Cooled Nuclear Power Plants
10:40 107 Lorenzo Antola, Andrea Cambriani, Katia Slavcheva, Fabrizio Trenta - Italy Nuclear Infrastructure and Institutional Capacity Restarting a Nuclear Program: the Italian Case
10:40 108 Igor Pioro, Sarah Mokry, M. Miletić, L. Grande, Eu. Saltanov, W. Peiman - Canada Generation IV NPPs: Supercritical “Steam” Rankine Thermodynamic Cycles Options 10:40 109 Sabina Markelj, Iztok Čadež, Zdravko Rupnik - Slovenia Production of Vibrationally Excited Hydrogen Molecules from Tungsten 10:40 110 Raphael Heffron, Kalev Kallemets - United Kingdom Estonia: The Prospects for Nuclear Energy Reactor Physics, Fuel Cycle and Research Reactors 10:40 201 Davor Grgić, Radomir Ječmenica, Dubravko Pevec - Croatia Spectral Codes Pin Power Prediction Comparison 10:40 205 Marjan Kromar, Bojan Kurinčič - Slovenia DRAGON and CORD-2 Nuclear Calculation of the NPP Krško Fuel Assembly
10:40 206 Gašper Žerovnik, Luka Snoj - Slovenia Evaluation of Integral Benchmark Experiments Uncertainty due to Boron Isotopic Abundance Variations
10:40 207 Dinka Vragolov, Mario Matijević, Dubravko Pevec - Croatia Modeling of Pool Critical Assembly Pressure Vessel Facility Benchmark 10:40 208 Ali Pazirandeh, Elham Younesian - Iran Linac Based Subcritical Th232-U233 Research Reactor 10:40 210 Gašper Žerovnik, Andrej Trkov, Christophe Destouche - Slovenia Extension of the MATSSF Code for Self-shielding Factor Calculations of Heterogeneous Samples
Reactor Operation 10:40 302 Slavko Slavič, Marjan Kromar, Bojan Žefran - Slovenia Latest Extensions of the FAR - Software Package for the Nuclear Material Accounting 10:40 303 Dražen Krajina, Dave Bubeck - USA Ovation® to Third Party Interface Solutions 10:40 305 Martin Kropík, Jan Rataj - Czech Republic New Control and Data Acquisition of Experiments at VR 1 Training Reactor Thermal Hydraulics 10:40 401 Ernest Hauser, Herb Estrada - USA The Impact on Flow Meter Uncertainty of Tubular Flow Conditioners 10:40 403 Ivo Kljenak, Borut Mavko - Slovenia Natural Circulation Simulation with Lumped-Parameter Codes Using Input Models Based on CFD Simulation
10:40 404 Aleksander Churbanov - Russia CFD Analysis of Free Surface Flows Using an Open Source Code 10:40 405 Ionut Anghel, Henryk Anglart, Stellan Hedberg - Sweden Experimental Investigation of Post-Dryout Heat Transfer in Annulus with Flow 10:40 406 François Henry, Yann Bartosiewitz - Belgium A Model for Droplet Entrainment in Horizontal Stratified Flow 10:40 407 Andrej Prošek - Slovenia Loss of External Load Analysis using RELAP5/MOD3.3 Patch 03 Computer Code 10:40 409 Tobias Zieger, Jay Faramarzi, Mark Hollerbach, Herbert Miller - Switzerland Controlling Valve Fluid Jet Energy to eliminate Vibration 10:40 410 Faramarz Yousefpour, Kaveh Karimi, Hamid Soltani - Iran Comparative Reliability Analysis for Two Different Designs of Residual Heat Removal System (RHRS) and Containment Spray System (CSS) of IR-360 Nuclear Power Plant
10:40 411 Davide Papini, Giovanni Pastore, Antonio Cammi - Italy Preliminary CFD Study of Flow Oscillations in Parallel Channels Using the Volume Of Fluid (VOF) Method
10:40 413 Marco Colombo, Antonio Cammi, Davide Papini, Marco Ricotti - Italy Numerical Investigation on Boiling Channel Instabilities by Imposing Constant Pressure Drop Boundary Condition Via a Large Bypass
10:40 414 Ilijana Iveković, Tomislav Bajs, Ivica Bašić - Croatia Screening Methodology for the Evaluation of PTS scenarios in the Symptom based Emergency Operating Procedures
10:40 415 Davor Grgić, Vesna Benčik, Siniša Šadek - Croatia Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation 10:40 416 Tomas Kačegavičius - Lithuania The Integral Analysis of 40 mm Diameter Pipe Rupture in Cooling System of Fusion Facility W7-X with ASTEC Code
10:40 417 Boštjan Končar, Carlos Estrada-Perez, Yassin Hassan - Slovenia Numerical Simulation of Turbulent Subcooled Boiling Flow in a Rectangular Channel Multiphyiscs 10:40 502 Peter Hermansky, Marian Krajčovič - Slovak Republic The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit
10:40 503 Dino Araneo, Giuseppe Agresta, Francesco D'Auria - Italy Fracture Mechanics Analysis in a Pressurized Heavy Water Reactor Vessel during LOCA Scenario
10:40 506 J. Juanas, T. Barrachina, R. Miró, R. Macián, G. Verdú - Spain Uncertainty and Sensitivity Aanlysis in the Neutronic Parameters Generation for BWR andPWR Coupled Thermalhydraulic -Neutronic Simulations Materials and Structural Integrity 10:40 601 Dumitra Lucan - Romania Corrosion Experience with the Secondary Side of CANDU Steam Generator Tubesheet 10:40 602 Igor Simonovski, Leon Cizelj - Slovenia Towards Modeling Intergranular Stress Corrosion Cracks on Grain Size Scales
10:40 603 Vladimír Slugeň, Jozef Lipka, Julius Dekan, Jarmila Degmova, Ignac Tóth, Pavol Šeliga, Ivan Smieško - Slovak Republic Observation of Long-term Corrosion at Nuclear Power Plant Bohunice (Slovakia) 10:40 605 Mihaela Uplaznik, Leon Cizelj, Igor Simonovski - Slovenia Some Consistency and Quality Tests for Finite Element Models of Grain and Grain Boundaries in Polycrystals
10:40 606 Jose M. Izquierdo, Javier Hortal, Miguel Sanchez, Enrique Meléndez, César Queral, Luisa Ibáñez, Israel Canamón, Ernesto Villalba, Jesus Gil, Ivan Fernández, Santiago Murcia, Javier Gómez - Spain Damage Domain Approach as a Strategy of Damage Exceedance Frequency Computation 10:40 608 Vitaly Surin, Nikolay Evstyukhin, Yury Kapralov, Anton Morozov - Russia High-effective Control System for Reactor Technological Equipment 10:40 609 Marjan Suban, Simon Božič, Andrej Zajec, Robert Cvelbar, Borut Bundara - Slovenia Analysis of Cracks Resulting from Thermite Welding of Cathodic Protection 10:40 611 Vitaly Surin, Nikolay Evstyukhin, Yury Kapralov, Evgeny Kapralov - Russia The Electronic Structure and Electrophysical Properties of Perspective Nuclear Fuel Probabilistic Safety Assessment 10:40 701 Faramarz Yousefpour, Davood Babazadeh, Hamid Soltani - Iran Sensitivity Analysis of Emergency Diesel Generator Availability in IR-360 Nuclear Power Plant Using Fault Tree Method
10:40 703 Antonio Fernandez, Paul Rizzo - USA Development of Probabilistic Seismic Hazard Analysis for Various International Locations; Challenges and Guidelines
10:40 704 Blaže Gjorgiev, Marko Čepin, Atanas Iliev - Slovenia Optimal Generation Schedule of Power System Considering Nuclear Power Plant with Application of Genetic Algorithms
10:40 705 Marko Čepin, Andrija Volkanovski - Slovenia Advantages and Difficulties With the Application of Methods of Probabilistic Safety Assessment to the Power Systems Reliability 10:40 706 Andrija Volkanovski - Slovenia Ageing Prioritization Based on PSA Results Severe Accidents 10:40 802 Milan Amižić, Siniša Šadek, Nenad Debrecin - Croatia In-vessel Severe Accident Analysis of PWR Station Blackout with the RELAP5/SCDAP and ASTEC Codes
10:40 803 Alexei Miassoedov, Xiaoyang Gaus-Liu, Thomas Jordan, L. Meyer, W. Tromm, Martin Steinbrück - Germany LACOMECO Experimental Platform at KIT 10:40 805 Fernando Robledo, Luis E. Herranz - Spain Current Activities Of CSN in the Field of Severe Accident Research 10:40 806 Venceslav Kostadinov - Slovenia Developing New Methodology for Nuclear Vulnerability Assessment 10:40 807 Matjaž Leskovar, Mitja Uršič - Slovenia Analysis of Melt Droplets Crust Growth During Steam Explosion Premixing Phase 10:40 808 Srinivasa Visweswaran, David Finnicum - USA Peer Review of Trial Application of Low Power and Shutdown PRA Standard 10:40 809 Emilie Beuzet, Jean Lamy, Hadrien Perron, Eric Simoni - France Modelling of Ruthenium Release in Air and Steam Atmospheres under Severe Accident Conditions using the MAAP4 Code
10:40 810 Mitja Uršič, Matjaž Leskovar, Borut Mavko - Slovenia Simulations of KROTOS Alumina and Corium Experiments: Applicability of the Improved Solidification Influence Modelling
10:40 811 Eugenio Coscarelli, Alessandro Del Nevo, Francesco D'Auria - Italy Numerical Investigation of Natural Circulation during a Small Break LOCA Scenarios in a PWR-System using the TRACE v5.0 code
10:40 812 Alexander Vasiliev - Russia Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V2 to Bottom Water Reflood Experiment PARAMETER-SF4
Radioactive Waste and Decommissioning 10:40 901 Nadja Železnik - Slovenia Perception of Radioactivity and Attitudes towards Radioactive Waste 10:40 904 Polona Tavčar, Igor Osojnik, Maks Pečnik - Slovenia Regulatory Experiences with the Unconditional Clearance of Radioactive Material in Slovenia
10:40 906 Ilija Plećaš, Slavko Dimovic - Serbia Leaching Study in Processs of Solidification of Radionuclide 54Mn in Concrete 10:40 907 Jan Smutek, Jiří Svoboda - Czech Republic Teaching Activities of Centre of Experimental Geotechnics Related to Radioactive Waste Storage Based on Research Experience Nuclear Fusion and Plasma Technology 10:40 1001 Martin Draksler, Boštjan Končar - Slovenia Heat Transfer and Jet Interaction for Different Arrays of Impinging Jets 10:40 1003 Rok Zaplotnik, Alenka Vesel, Miran Mozetič - Slovenia Development of a Large Plasma Reactor for Removal of Deposits in Fusion Reactors 10:40 1004 Alenka Vesel, Miran Mozetič, Marianne Balat - Pichelin, Rok Zaplotnik - Slovenia Erosion of W-C composite in Hydrogen Plasma at Temperatures Above 1000 K 10:40 1005 Melita Lenošek, Saša Novak - Slovenia Travelling Exhibition – Fusion Expo 10:40 1006 Tea Toplišek, Goran Dražić, Spomenka Kobe, Vilibald Bukošek - Slovenia Microstructure and Mechanical Properties of SiC Fibers for Potentional Use in a Future Fusion Reactor
10:40 1007 Luka Snoj, Brian Syme, Sergey Popovichev - Slovenia Calculations to Support JET Neutron Yield Calibration: Neutron Scattering in Source Holder
10:40 1008 Jernej Kovačič, Tomaž Gyergyek, Milan Čerček - Slovenia A Fully-kinetic PIC Simulation of an Emissive Probe in Tokamak Relevant Plasma 10:40 1009 Vida Žigman - Slovenia Vibrationally Excited Hydrogen Molecules from Desorptive Recombination at Surfaces 10:40 1011 Janez Krek, Nikola Jelić, Jože Duhovnik - Slovenia Particle-In-Cell (PIC) Simulations and Grid-Free Treecode Method 10:40 1012 Primož Pelicon, Primož Vavpetič, Zdravko Rupnik, Iztok Čadež, Sabina Markelj, Nataša Grlj, Mirko Ribič, Zvone Grabnar - Slovenia Focused 3He Ion Beam: Highly Selective and Laterally Resolving Method for Deuterium Detection in Plasma-facing Components
10:40 1013 Bojan Zajec, Vincenc Nemanič, C. Porosnicu, C. Lungu - Slovenia Hydrogen Permeability of Beryllium Films 10:40 1015 Alberto Milocco, Andrej Trkov, M. Pillon, Roberto Bedogni - Slovenia A Monte Carlo Model of the D-D Neutron Source at FNG 10:40 1016 Jernej Kovačič, Tomaž Gyergyek, Milan Čerček - Slovenia A Fluid Model and PIC Simulation of Two-Electron-Temperature Plasma in an Oblique Magnetic Field
10:40 1017 M. Moosavi, Abbas Ghasemizad - Iran The Investigation of Fuel Energy Gain for Tritium-Poor Fuels in Fast Ignition Fusion 10:40 1018 Leon Kos, Jože Duhovnik, Nikola Jelić - Slovenia The Shape of the Potential Profile Near the Boundary in the Tonks-Langmuir Model to the Case of Finite Ion–source Temperature
10:40 1019 Aljaž Ivekovič, Saša Novak, Goran Dražić - Slovenia SITE-P: A Novel Route for Preparation of SiCf/SiC Composites 10:40 1020 Goran Dražić, Tea Toplišek, Aljaž Ivekovič, Saša Novak - Slovenia Compatibility of W-core β-SiC Fibers with SiC based Composite Material for Fusion Application Radiation and Environment Protection 10:40 1101 Milko Križman, Michel Cindro, Barbara Vokal Nemec - Slovenia Discharged Radionuclides from the Slovenian Nuclear & Radiation Facilities and Their Actual Detection in the Environment
10:40 1102 Thomas Breznik, Marko Gerbec, Borut Smodiš - Slovenia Task Analysis and Risk Assessment in Case of Accident Involving Transport of Radioactive Materials by Road in the Republic of Slovenia Education, Public Relations and Regulatory Issues 10:40 1202 Radko Istenič, Igor Jenčič - Slovenia Public Opinion about Nuclear Energy – Year 2010 Poll 10:40 1203 David Helling, David Kwiatkowski, Nathan Hall, Pamela Aigner - USA-Pensylvania Adopting Curriculum Integration to Improve Nuclear Training and Education 10:40 1204 Tatjana Frelih Kovačič, Janez Češarek, Igor Osojnik, Maks Pečnik - Slovenia Implementation of the Code of Conduct on the Safety and Security of Radioactive Sources in Slovenia
10:40 1205 Evgeny Kapralov, Yury Kapralov, Gennady Filimonov - Russia Training of Specialists in the Assessment of NPP Equipment Compliance 10:40 1207 Tea Bilić-Zabric - Slovenia Deployment of New Nuclear Power Plant in Slovenia 10:40 1208 Veronika Simonovska, Ulrik Von Estorff - Netherlands EHRO-N: a Tool Complementing Instruments and Initiatives for Improved Management of Nuclear Human Resources in the European Union
10:40 1209 Igor Fifnja, Matjaž Žvar, Dušan Češnjevar - Slovenia Krško NPP Full Scope Simulator Utilization Experience
Session 10 11:40 Reactor Operation
11:40 304 Martin Chambers, Dejvi Kadivnik, Bojan Kurinčič - Slovenia Investigation of Grid-To-Rod-Fretting at Krško Nuclear Power Plant 12:00 301 Peter Schimann, Zoran V. Stošić - Germany Operational Know-How And Know-Why – AREVA´s Internal And External Exchange Of Information And Experience
Session 7 14:00 Invited Lecture 5
14:00 4 Reko Rantamäki - Finland Nuclear Renaissance? Case Loviisa 3, Finland
Session 8 14:40 Nuclear Renaissance and Prospects of Nuclear Energy
14:40 111 Rolando Calabrese - Italy Investigation in the Near and Long Term Perspective of Italian Scenario 15:00 103 Jurij Avsec, Peter Virtič, Andrej Predin, Luka Štrubelj, Tomaž Žagar - Slovenia Economy Analysis of Electricity Production from Hydrogen in Combination with Nuclear Power Plant
15:20 104 Paolo Gaio - Belgium AP1000 on Schedule for 2013: Status of Construction in China and USA
Session 18 16:00 Education and Public Relations
16:00 1210 Oscar Cabellos de Francisco, C. Ahnert, D. Cuervo, N. Garcia Herranz, E. Gallego, E. Minguez, J.M. Aragones, A. Lorente, A. Piedra - Spain Education and Training of Future Nuclear Engineers at DIN: From Advanced Computer Codes to Interactive Plant Simulator
16:20 1206 Siniša Cimeša, Andreja Peršič, Leopold Vrankar, Andrej Stritar - Slovenia Analysis of Human Resources and Technical Knowledge needed for the Licensing of the
New Nuclear Build: the SNSA Approach
16:40 1201 Michel Giot - Belgium A Survey of the Contribution of Large Experimental Nuclear Facilities to Education in Europe
Session 19 17:20 National Energy Programme and NPP Krsko 2 Project Status
Nuclear Renaissance and Prospects of Nuclear Energy 17:20 101 Andreja Urbančič, Boris Sučič - Slovenia Challenges of the New Slovenian Energy Program Proposal
_Table of Contents
Invited Lectures
The Role of Nuclear Data for Fusion Technology Studies Robin Forrest 2 Models and Their Benchmarking - International Experience in Nuclear Applications Enrico Sartori 2 Fusion Yield Measurements on JET and their Calibration Brian Syme 3 Nuclear Renaissance? Case Loviisa 3, Finland Reko Rantamäki 4 Cross-Cutting Tasks on Thermal-Hydraulics of Innovative Nuclear Systems Cheng Xu
Nuclear Renaissance and Prospects of Nuclear Energy
Challenges of the New Slovenian Energy Program Proposal Andreja Urbančič, Boris Sučič 6 A Multi-Physics Numerical Model for the MSR Core Dynamics Claudia Guerrieri, Antonio Cammi, Carlo Fiorina, Lelio Luzzi 7 Economy Analysis of Electricity Production from Hydrogen in Combination with Nuclear Power Plant Jurij Avsec, Peter Virtič, Andrej Predin, Luka Štrubelj, Tomaž Žagar 7 AP1000 on Schedule for 2013: Status of Construction in China and USA Paolo Gaio 8 Development of a Simulation Tool for a Preliminary Analysis of the MSR Core Dynamics Carlo Fiorina, Antonio Cammi, Claudia Guerrieri, Lelio Luzzi, Benedetto Spinelli 9 Appropriate Thermodynamic Cycles to be Used in Future Pressure-Channel Supercritical Water-Cooled Nuclear Power Plants Laure Lizon-A-Lugrin, Alberto Teyssedou, Igor Pioro 10 Nuclear Infrastructure and Institutional Capacity Restarting a Nuclear Program: the Italian Case Lorenzo Antola, Andrea Cambriani, Katia Slavcheva, Fabrizio Trenta 11 Generation IV NPPs: Supercritical “Steam” Rankine Thermodynamic Cycles Options Igor Pioro, Sarah Mokry, M. Miletić, L. Grande, Eu. Saltanov, W. Peiman 12 Estonia: The Prospects for Nuclear Energy Raphael Heffron, Kalev Kallemets 13 Investigation in the Near and Long Term Perspective of Italian Scenario Rolando Calabrese 14
Reactor Physics, Fuel Cycle and Research Reactors
Spectral Codes Pin Power Prediction Comparison Davor Grgić, Radomir Ječmenica, Dubravko Pevec 16 Building a Dynamic Monte Carlo Code to Simulate New Reactor Concepts Nikolas Catsaros, B. Gaveau, M. Jaekel, Jacques Maillard, G. Maurel, P. Savva, J. Silva, 16 M. Varvayanni Neutronics Analysis of the TRIGA Vienna Mixed Core Helmuth Böck, Rustam Khan 17 Methodology Investigations on Uncertainties Propagation in Nuclear Data Evaluation P. Dossantos-Uzarralde, H.P. Jacquet, G. Gauriot 17 DRAGON and CORD-2 Nuclear Calculation of the NPP Krško Fuel Assembly Marjan Kromar, Bojan Kurinčič 18 Evaluation of Integral Benchmark Experiments Uncertainty due to Boron Isotopic Abundance Variations Gašper Žerovnik, Luka Snoj 18 Modeling of Pool Critical Assembly Pressure Vessel Facility Benchmark Dinka Vragolov, Mario Matijević, Dubravko Pevec 19 Linac Based Subcritical Th232-U233 Research Reactor Ali Pazirandeh, Elham Younesian 20 MOX Benchmark Calculation Results by Monte Carlo and Deterministic Codes Marco Pecchia, G. Kotev, Carlo Parisi, Francesco D'Auria 21 Extension of the MATSSF Code for Self-shielding Factor Calculations of Heterogeneous Samples Gašper Žerovnik, Andrej Trkov, Christophe Destouche 21
Reactor Operation
Operational Know-How And Know-Why – AREVA´s Internal And External Exchange Of Information And Experience Peter Schimann, Zoran V. Stošić 24 Latest Extensions of the FAR - Software Package for the Nuclear Material Accounting Slavko Slavič, Marjan Kromar, Bojan Žefran 25 Ovation® to Third Party Interface Solutions Dražen Krajina, Dave Bubeck 25 Investigation of Grid-To-Rod-Fretting at Krško Nuclear Power Plant Martin Chambers, Dejvi Kadivnik, Bojan Kurinčič 26 New Control and Data Acquisition of Experiments at VR 1 Training Reactor Martin Kropík, Jan Rataj 27
Thermal Hydraulics
The Impact on Flow Meter Uncertainty of Tubular Flow Conditioners Ernest Hauser, Herb Estrada 30 Steam Bubble Condensation in Polydispersed Flow –Experiments and CFD Simulations Eckhard Krepper, Martin Schmidtke, Dirk Lucas, Matthias Beyer 31 Natural Circulation Simulation with Lumped-Parameter Codes Using Input Models Based on CFD Simulation Ivo Kljenak, Borut Mavko 32 CFD Analysis of Free Surface Flows Using an Open Source Code Aleksander Churbanov 33 Experimental Investigation of Post-Dryout Heat Transfer in Annulus with Flow Obstacles Ionut Anghel, Henryk Anglart, Stellan Hedberg 34 A Model for Droplet Entrainment in Horizontal Stratified Flow François Henry, Yann Bartosiewitz 35 Loss of External Load Analysis using RELAP5/MOD3.3 Patch 03 Computer Code Andrej Prošek 35 Direct Numerical Simulation of Heat Transfer at Very Low Prandtl Number: Application to convection convection in Lead-Bismuth flows Laurent Bricteux, Matthieu Duponcheel, Yann Bartosiewitz 36 Controlling Valve Fluid Jet Energy to eliminate Vibration Tobias Zieger, Jay Faramarzi, Mark Hollerbach, Herbert Miller 37 Comparative Reliability Analysis for Two Different Designs of Residual Heat Removal System (RHRS) and Containment Spray System (CSS) of IR-360 Nuclear Power Plant Faramarz Yousefpour, Kaveh Karimi, Hamid Soltani 38 Preliminary CFD Study of Flow Oscillations in Parallel Channels Using the Volume Of Fluid (VOF) Method Davide Papini, Giovanni Pastore, Antonio Cammi 39 Analysis of the Two-phase Flow Phenomena at PGV Test Facility Yuriy Parfenov, Vladimir Melikhov, Oleg Melikhov, Alexey Nerovnov 40 Numerical Investigation on Boiling Channel Instabilities by Imposing Constant Pressure Drop Boundary Condition Via a Large Bypass Marco Colombo, Antonio Cammi, Davide Papini, Marco Ricotti 40 Screening Methodology for the Evaluation of PTS scenarios in the Symptom based Emergency Operating Procedures Ilijana Iveković, Tomislav Bajs, Ivica Bašić 41 Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation Davor Grgić, Vesna Benčik, Siniša Šadek 42 The Integral Analysis of 40 mm Diameter Pipe Rupture in Cooling System of Fusion Facility W7-X with ASTEC Code Tomas Kačegavičius 42 Numerical Simulation of Turbulent Subcooled Boiling Flow in a Rectangular Channel Boštjan Končar, Carlos Estrada-Perez, Yassin Hassan 43
Multiphyiscs
Implementation of New Simulation Capabilities in RELAP5/PARCS v2.7 Coupled Codes R. Miró, C. Pereira, T. Barrachina, G. Verdú, J.C. Martínez-Murillo 46 The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit Peter Hermansky, Marian Krajčovič 47 Fracture Mechanics Analysis in a Pressurized Heavy Water Reactor Vessel during LOCA Scenario Dino Araneo, Giuseppe Agresta, Francesco D'Auria 48 TRANSURANUS Verification Against CNEA’s PHWR MOX Experiments, from IFPE Database Davide Rozzia, Martina Adorni, Alessandro Del Nevo, Francesco D'Auria 49 Reactivity Initiated Accident Analysis of the HPLWR Three Pass Core with Ascending Gap Flow Using the KIKO3D-ATHLET Code György Hegyi, András Keresztúri, Csaba Maŕaczy, Istvan Trosztel 50 Uncertainty and Sensitivity Aanlysis in the Neutronic Parameters Generation for BWR andPWR Coupled Thermalhydraulic -Neutronic Simulations J. Juanas, T. Barrachina, R. Miró, R. Macián, G. Verdú 51
Materials and Structural Integrity
Corrosion Experience with the Secondary Side of CANDU Steam Generator Tubesheet Dumitra Lucan 53 Towards Modeling Intergranular Stress Corrosion Cracks on Grain Size Scales Igor Simonovski, Leon Cizelj 53 Observation of Long-term Corrosion at Nuclear Power Plant Bohunice (Slovakia) Vladimír Slugeň, Jozef Lipka, Julius Dekan, Jarmila Degmova, Ignac Tóth, Pavol Šeliga, 54 Ivan Smieško Modular 3-D Solid Finite Element Model for Fatigue Analyses of a PWR Reactor Coolant System Oriol Garrido, Leon Cizelj, Igor Simonovski 54 Some Consistency and Quality Tests for Finite Element Models of Grain and Grain Boundaries in Polycrystals Mihaela Uplaznik, Leon Cizelj, Igor Simonovski 55 Damage Domain Approach as a Strategy of Damage Exceedance Frequency Computation Jose M. Izquierdo, Javier Hortal, Miguel Sanchez, Enrique Meléndez, César Queral, Luisa 56 Ibáñez, Israel Canamón, Ernesto Villalba, Jesus Gil, Ivan Fernández, Santiago Murcia, Javier Gómez Slovenian Regulatory Approach to Design Lifetime Extension Artur Muehleisen, Matjaž Podjavoršek, Andreja Peršič, Djordje Vojnovič, Andrej Stritar 56 High-effective Control System for Reactor Technological Equipment Vitaly Surin, Nikolay Evstyukhin, Yury Kapralov, Anton Morozov 57 Analysis of Cracks Resulting from Thermite Welding of Cathodic Protection Marjan Suban, Simon Božič, Andrej Zajec, Robert Cvelbar, Borut Bundara 58 AREVA’s Fatigue Concept (AFC) – An Integrated And Multidisciplinary Approach To The Fatigue Assessment Of NPP Components Christian Poeckl, Steffen Bergholz, Juergen Rudolph, Nikolaus Wirtz 58 The Electronic Structure and Electrophysical Properties of Perspective Nuclear Fuel Vitaly Surin, Nikolay Evstyukhin, Yury Kapralov, Evgeny Kapralov 59 Probabilistic Safety Assessment
Sensitivity Analysis of Emergency Diesel Generator Availability in IR-360 Nuclear Power Plant Using Fault Tree Method Faramarz Yousefpour, Davood Babazadeh, Hamid Soltani 61 Sensitivity and Uncertainty Analysis for Age-Dependent Model of Test and Maintenance Duško Kančev, Marko Čepin 61 Development of Probabilistic Seismic Hazard Analysis for Various International Locations; Challenges and Guidelines Antonio Fernandez, Paul Rizzo 62 Optimal Generation Schedule of Power System Considering Nuclear Power Plant with Application of Genetic Algorithms Blaže Gjorgiev, Marko Čepin, Atanas Iliev 63 Advantages and Difficulties With the Application of Methods of Probabilistic Safety Assessment to the Power Systems Reliability Marko Čepin, Andrija Volkanovski 64 Ageing Prioritization Based on PSA Results Andrija Volkanovski 64
Severe Accidents
Implementation of Severe Accident Management Guidelines to Shutdown and Low-Power Operation Modes for VVER and PWR Plants Oleg Solovjanov, R. Lutz, Nathalie Dessars, Thibaut Rensonnet 67 In-vessel Severe Accident Analysis of PWR Station Blackout with the RELAP5/SCDAP and ASTEC Codes Milan Amižić, Siniša Šadek, Nenad Debrecin 68 LACOMECO Experimental Platform at KIT Alexei Miassoedov, Xiaoyang Gaus-Liu, Thomas Jordan, L. Meyer, W. Tromm, Martin 68 Steinbrück Study of Boron Behaviour in the Primary Circuit of Water Reactors under Severe Accident Conditions; a Comparison of Recent Integral and Separate-Effect Data Tim Haste, Frédéric Payot, Cristina Dominguez, Philippe March, B. Simondi-Teisseire, 70 Martin Steinbrück, Roland Zeyen Current Activities Of CSN in the Field of Severe Accident Research Fernando Robledo, Luis E. Herranz 71 Developing New Methodology for Nuclear Vulnerability Assessment Venceslav Kostadinov 72 Analysis of Melt Droplets Crust Growth During Steam Explosion Premixing Phase Matjaž Leskovar, Mitja Uršič 72 Peer Review of Trial Application of Low Power and Shutdown PRA Standard Srinivasa Visweswaran, David Finnicum 73 Modelling of Ruthenium Release in Air and Steam Atmospheres under Severe Accident Conditions using the MAAP4 Code Emilie Beuzet, Jean Lamy, Hadrien Perron, Eric Simoni 75 Simulations of KROTOS Alumina and Corium Experiments: Applicability of the Improved Solidification Influence Modelling Mitja Uršič, Matjaž Leskovar, Borut Mavko 76 Numerical Investigation of Natural Circulation during a Small Break LOCA Scenarios in a PWR-System using the TRACE v5.0 code Eugenio Coscarelli, Alessandro Del Nevo, Francesco D'Auria 76 Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V2 to Bottom Water Reflood Experiment PARAMETER-SF4 Alexander Vasiliev 77
Radioactive Waste and Decommissioning
Perception of Radioactivity and Attitudes towards Radioactive Waste Nadja Železnik 79 Preliminary Design of Vrbina LILW Repository Boštjan Duhovnik, Janja Špiler 80 Transport of Radioactive Waste from Small Producers Marko Kostanjevec, Marija Fabjan, Simona Sučić 81 Regulatory Experiences with the Unconditional Clearance of Radioactive Material in Slovenia Polona Tavčar, Igor Osojnik, Maks Pečnik 81 EU Law on Radioactive Waste Ana Stanič 82 Leaching Study in Processs of Solidification of Radionuclide 54Mn in Concrete Ilija Plećaš, Slavko Dimovic 83 Teaching Activities of Centre of Experimental Geotechnics Related to Radioactive Waste Storage Based on Research Experience Jan Smutek, Jiří Svoboda 83 Revision 2 of the Program of NPP Krško Decommissioning and SF & LILW Disposal Nadja Železnik, Metka Kralj, Irena Mele, Primož Stropnik, Ivica Levanat, Vladimir 84 Lokner, Andrea Rapić
Nuclear Fusion and Plasma Technology
Heat Transfer and Jet Interaction for Different Arrays of Impinging Jets Martin Draksler, Boštjan Končar 87 Calculations to Support JET Neutron Yield Calibration: Contributions to the External Neutron Monitor Responses Luka Snoj, Brian Syme, Sergey Popovichev, Igor Lengar, Sean Conroy 88 Development of a Large Plasma Reactor for Removal of Deposits in Fusion Reactors Rok Zaplotnik, Alenka Vesel, Miran Mozetič 89 Erosion of W-C composite in Hydrogen Plasma at Temperatures Above 1000 K Alenka Vesel, Miran Mozetič, Marianne Balat - Pichelin, Rok Zaplotnik 89 Travelling Exhibition – Fusion Expo Melita Lenošek, Saša Novak 90 Microstructure and Mechanical Properties of SiC Fibers for Potentional Use in a Future Fusion Reactor Tea Toplišek, Goran Dražić, Spomenka Kobe, Vilibald Bukošek 91 Calculations to Support JET Neutron Yield Calibration: Neutron Scattering in Source Holder Luka Snoj, Brian Syme, Sergey Popovichev 92 A Fully-kinetic PIC Simulation of an Emissive Probe in Tokamak Relevant Plasma Jernej Kovačič, Tomaž Gyergyek, Milan Čerček 93 Vibrationally Excited Hydrogen Molecules from Desorptive Recombination at Surfaces Vida Žigman 94 ITER, an Essential Step in Fusion Development Jesus Izquierdo 94 Particle-In-Cell (PIC) Simulations and Grid-Free Treecode Method Janez Krek, Nikola Jelić, Jože Duhovnik 95 Focused 3He Ion Beam: Highly Selective and Laterally Resolving Method for Deuterium Detection in Plasma-facing Components Primož Pelicon, Primož Vavpetič, Zdravko Rupnik, Iztok Čadež, Sabina Markelj, Nataša 96 Grlj, Mirko Ribič, Zvone Grabnar Hydrogen Permeability of Beryllium Films Bojan Zajec, Vincenc Nemanič, C. Porosnicu, C. Lungu 97
Mechanical Properties and Microstructural Characterizations of Potassium Doped Tungsten Hua Sheng, Inge Uytdenhouwen, Vincent Massaut, Guido Van Oost, Jozef Vleugels 98 A Monte Carlo Model of the D-D Neutron Source at FNG Alberto Milocco, Andrej Trkov, M. Pillon, Roberto Bedogni 99 A Fluid Model and PIC Simulation of Two-Electron-Temperature Plasma in an Oblique Magnetic Field Jernej Kovačič, Tomaž Gyergyek, Milan Čerček 99 The Investigation of Fuel Energy Gain for Tritium-Poor Fuels in Fast Ignition Fusion M. Moosavi, Abbas Ghasemizad 100 The Shape of the Potential Profile Near the Boundary in the Tonks-Langmuir Model to the Case of Finite Ion–source Temperature Leon Kos, Jože Duhovnik, Nikola Jelić 101 SITE-P: A Novel Route for Preparation of SiCf/SiC Composites Aljaž Ivekovič, Saša Novak, Goran Dražić 102 Compatibility of W-core β-SiC Fibers with SiC based Composite Material for Fusion Application Goran Dražić, Tea Toplišek, Aljaž Ivekovič, Saša Novak 103 Production of Vibrationally Excited Hydrogen Molecules from Tungsten Sabina Markelj, Iztok Čadež, Zdravko Rupnik 104
Radiation and Environment Protection
Discharged Radionuclides from the Slovenian Nuclear & Radiation Facilities and Their Actual Detection in the Environment Milko Križman, Michel Cindro, Barbara Vokal Nemec 106 Task Analysis and Risk Assessment in Case of Accident Involving Transport of Radioactive Materials by Road in the Republic of Slovenia Thomas Breznik, Marko Gerbec, Borut Smodiš 106 Comparison of Measured Activity Concentrations in Plants From the Area of the Former Uranium Mine at Žirovski Vrh, Slovenia with Activity Concentrations Obtained by the ERICA Biota Assessment Tool Marko Černe, Borut Smodiš 107 A Brief History of Krško NPP Radiation Impact on Environment Matjaž Koželj 108
Education, Public Relations and Regulatory Issues
A Survey of the Contribution of Large Experimental Nuclear Facilities to Education in Europe Michel Giot 110 Public Opinion about Nuclear Energy – Year 2010 Poll Radko Istenič, Igor Jenčič 111 Adopting Curriculum Integration to Improve Nuclear Training and Education David Helling, David Kwiatkowski, Nathan Hall, Pamela Aigner 111 Implementation of the Code of Conduct on the Safety and Security of Radioactive Sources in Slovenia Tatjana Frelih Kovačič, Janez Češarek, Igor Osojnik, Maks Pečnik 112 Training of Specialists in the Assessment of NPP Equipment Compliance Evgeny Kapralov, Yury Kapralov, Gennady Filimonov 113 Analysis of Human Resources and Technical Knowledge needed for the Licensing of the New Nuclear Build: the SNSA Approach Siniša Cimeša, Andreja Peršič, Leopold Vrankar, Andrej Stritar 114 Deployment of New Nuclear Power Plant in Slovenia Tea Bilić-Zabric 114 EHRO-N: a Tool Complementing Instruments and Initiatives for Improved Management of Nuclear Human Resources in the European Union Veronika Simonovska, Ulrik Von Estorff 115 Krško NPP Full Scope Simulator Utilization Experience Igor Fifnja, Matjaž Žvar, Dušan Češnjevar 116 Education and Training of Future Nuclear Engineers at DIN: From Advanced Computer Codes to Interactive Plant Simulator Oscar Cabellos de Francisco, C. Ahnert, D. Cuervo, N. Garcia Herranz, E. Gallego, E. 117 Minguez, J.M. Aragones, A. Lorente, A. Piedra Nuclear Energy f0r New Eur0pe 2010
Invited Lectures
1 Nuclear Energy f0r New Eur0pe 2010
_1 ______
The Role of Nuclear Data for Fusion Technology Studies
Robin A Forrest International Atomic Energy Agency Vienna International Centre PO Box 100 1400 Vienna, Austria
Nuclear data are of fundamental importance in studies of nuclear technology. In these studies are included experiments to measure cross sections or decay properties and simulations of the design of fission power plants, fusion devices and accelerators. The large amount of data required are stored in computer readable formats in data libraries and the most common of these are the general purpose files used for neutronics or transport calculations. These files also contain the standards against which most measurements are made. The other class of libraries are the special purpose ones containing decay data, fission yields and cross section data for dosimetry and activation.
This paper gives examples of what data are available and describes their use for various applications. Focus will be made on neutron-induced activation data with examples of how the reactions of particular importance can be identified. All data should be accompanied by estimates of the uncertainty. This is best achieved by including covariance data, however this is extremely challenging and only a subset of the available data have such uncertainty data. The general principles of how covariance matrices are used and possible methods of avoiding them are outlined.
_2 ______
Models and their Benchmarking - International Experience in Nuclear Applications
Enrico Sartori NEA Data Bank OECD, 12 bd des Iles, F- 92130, Issy-Les-Moulineaux, France
After the introduction of the concept of a model, the benchmarking procedure and purpose is presented with successful examples from thefields of reactor physics, criticality, and radiation shielding.
2 Nuclear Energy f0r New Eur0pe 2010
_3 ______
Fusion Yield Measurements on JET and their Calibration
D.B. Syme, S. Popovichev EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, UK S. Conroy EURATOM-VR Fusion Association, Uppsala University, SE-75120 Uppsala, Sweden I. Lengar, L. Snoj EURATOM-MHEST Slovenian Fusion Association, Ljubljana, Slovenija and JET EFDA contributors See the Appendix of F. Romanelli et al., Proceedings of the 22nd IAEA
The power output of fusion experiments and fusion reactor-like devices is measured in terms of the neutron emission rates which relate directly to the fusion yield rate. The largest fusion power produced in magnetically confined experiments so far was at JET in 1997, when a peak value of 16 MW was maintained for 0.5 second.
After introducing JET itself, the devices and methods measuring the neutron yields are described and the earlier calibration of JET is explained. This acts as an introduction to the planned new in-situ calibration for December 2010 which has to accommodate the many significant changes in JET itself since the last calibration.
A new, more detailed, calibration is being provided in practical terms by means of an engineering programme of development of the robotic tools which will allow safe & accurate deployment of a strong 252Cf source for the measurements. It is led by a scientific programme which seeks to better understand the limitations of the calibration, to optimise the measurements and other provisions, to provide corrections for perturbing factors and to ensure personnel safety and safe working conditions. Much of this work is based on an extensive programmed of Monte-Carlo calculations, including the updating of previous JET models to provide continuity of comparison with previous understanding, the provision of fast models for side effect estimation and the development of a new more detailed JET model in order to allow comparisons with the older more homogeneous model while coping with the demands of the new calibration.
This work has been conducted under EFDA and is partly funded by UK EPSRC and Euratom.
3 Nuclear Energy f0r New Eur0pe 2010
_4 ______
Nuclear Renaissance? Case Loviisa 3, Finland
Reko Rantamäki Fortum Power and Heat Oy PL 100, FI-00048 FORTUM, Finland [email protected]
After the completion of Fortum Power and Heat Oy's (Fortum) and Teollisuuden Voima Oy's (TVO) nuclear power plant units (Loviisa 1 and 2, Olkiluoto 1 and 2) at the beginning of 1980s there have been several projects in Finland that have been aimed for building new nuclear power plants. During the last decades there have been four attempts in which also Loviisa 3 has been involved. In 1986 an application went down because of Chernobyl. In 1993 an application was rejected by the parliament. In 2002 the parliament favoured TVO's Olkiluoto 3 application. The latest round started in spring 2007 when both Fortum and TVO announced that they are going to start an environmental impact assessment for new nuclear power plant units (Loviisa 3, Olkiluoto 4). Also in 2007, a new company called Fennovoima (major owner E.ON) emerged and announced that they are looking for a new nuclear power plant site, where they want to build a nuclear power plant. Each of these companies filed decision in principle applications after the environmental impact assessments were completed. In addition Posiva Oy, which takes care of its owners (TVO and Fortum) spent fuel final disposal filed two decision in principle applications, one for Olkiluoto 4 and one for Loviisa 3 spent fuel final disposal. All applications were considered together in the Council of State. Applications of TVO and Fennovoima were favoured by the Council of State while Fortum's Loviisa 3 application was rejected. Parliament ratified TVO's, Fennovoima's and Posiva's (Olkiluoto 4 spent fuel) favourable decisions in principle 1st of July 2010.
4 Nuclear Energy f0r New Eur0pe 2010
Nuclear Renaissance and Prospects of Nuclear Energy
5 Nuclear Energy f0r New Eur0pe 2010
_101 ______
Challenges of the New Slovenian Energy Program Proposal
Andreja Urbančič, Boris Sučić “Jožef Stefan” Institute, Energy Efficiency Centre Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
In Europe, although one of the most energy efficient regions in the world, further energy efficiency gains can significantly boost competitiveness. At least 20% of improvement by 2020 could be profitable and in that sense Slovenia is not an exemption. The new Slovenian Energy Program proposal has been developed having in mind four basic objectives – increasing reliability of energy supply, boosting competitiveness of the economy, promoting environmentally sustainable solutions and securing social cohesiveness. In the frame of the EU energy climate package Slovenia set ambitious target by 2020 to achieve 25% renewable energy in gross final energy demand. Also, its vision is that Slovenia would play an active role in the development and promotion of the new technologies for energy efficiency, renewable energy solutions in industry, residential and transport sector. For further development of the centralized electricity generation the program will be focused on competitiveness of producers in energy market, reliable energy mix, adequate capacities and the best available technologies, (BAT). The New Slovenian Energy Program will give the basis and will be an important step in decision making process for long term continuation of electricity production in nuclear power plants in Slovenia, as new unit in a range 1000 – 1600 MW is under consideration.
Enclosed with the new Slovenian Energy Program proposal is the impact assessments including among others costs of frameworks of measures from the perspectives of energy consumers, the general factors governing public finances and the development of the Slovenian energy sector and the economy. Special attention is given to the most influencing external factors for the energy sector development – prices in the international energy markets and the velocity of technology changes mainly in transport, electricity distribution and renewable energy sector.
With the imperative to keep this Program alive, Slovenian Ministry of Economy will continuously monitor its implementation. Government will have the responsibility to report and discuss all implementation issues annually in the Slovenian Parliament. Program has to be updated every five years. Only by this adaptive approach in policy and program evaluation and implementation of all the real-world aspects will be taken into account, which should ensure the full applicability of the new policy and program.
Within this paper evaluation of the most important challenges for the future development of the energy sector in Slovenia is presented. Special attention is given to the efficiency within the energy sector or how to make entire energy sector. Main elements of new energy program impact assessment are presented.
6 Nuclear Energy f0r New Eur0pe 2010
_102 ______
A Multi-Physics Numerical Model for the MSR Core Dynamics
Claudia Guerrieri, Antonio Cammi, Carlo Fiorina, Lelio Luzzi Politecnico di Milano – Department of Energy Nuclear Division-CeSNEF Via La Masa, 34 – 20156 Milano, Italy [email protected]
In the framework of the Generation IV International Forum (GIF), six innovative concepts of nuclear reactors have been chosen as suitable for the feature challenges of nuclear energy. Among these reactors, a renewed interest has been focused recently on the Molten Salt Reactor (MSR) due to its unique capabilities and competitive economics for actinide burning and extending fuel resources (Generation IV International Forum – 2008 annual report).
In MSRs the molten salt serves both as fuel and coolant, leading to a particular and complex physical environment. The dynamic study of such system requires a careful set-up of dedicated numerical models, able to account for the intrinsic coupling between neutronics and fluid dynamics. To this purpose, a multi-physics model has been developed for the study of the graphite-moderated MSR core dynamics. In this model, the partial differential equations governing the different phenomena are solved simultaneously in the same computational environment. As concerns the geometry, a single channel representative of the core average conditions has been modelled considering a two- dimensional axial-symmetric domain.
The mentioned model has been compared with a dedicated lumped model developed during the sixties at Oak Ridge National Laboratory for the MSRE (Molten Salt Reactor Experiment). This reactor has been chosen as reference because of the noteworthy availability of experimental data. Such comparison has lead to satisfactory results as concerns the general behaviour of the core. In addition, the developed multi-physics model provided careful estimates of important features such as the spatial distribution of neutron precursors and the temperature field in both the molten salt and the graphite.
_103 ______
Economy Analysis of Electricity Production from Hydrogen in Combination with Nuclear Power Plant
Jurij Avsec, Peter Virtič, Andrej Predin 1 University of Maribor, Faculty of Energy Technology Hočevarjev trg 1 8270 Krško, Slovenia. E-mail: [email protected]
Luka Štrubelj, Tomaž Žagar GEN energija d.o.o. Cesta 4. Julija 42 8270 Krško, Slovenia [email protected], [email protected]
Unlike fossil fuels, hydrogen is a clean energy carrier that does not react to produce greenhouse gases. Some have questioned whether the “hydrogen economy” is for us, or far in the future for our
7 Nuclear Energy f0r New Eur0pe 2010 grandkids. But the worldwide hydrogen market is already valued at over $300 billion/year, growing at over 10%/year, and expected to reach several $trillions/year by 2020. Hydrogen is used widely by petrochemical, agricultural (ammonia for fertilizers), manufacturing, food processing, electronics, plastics, metallurgical, aerospace and other industries. The usage in the transportation field will present significant market for hydrogen in the near future.
Efficient and sustainable methods of clean energy production, transmition and usage are needed in all countries of the world, having in mind depleting oil reserves and reduction of the carbon dioxide emissions. Nuclear energy as clean source of energy and hydrogen as a clean energy vector seems a good combination. In the paper we will show which technology is the most suitable for the hydrogen production in combination with nuclear power plant of generation II and III, which are currently producing electricity or are under construction. Some of the generation IV nuclear power plants can use other, more efficient technologies to produce hydrogen due to higher temperatures achieved at the core outlet.
The hydrogen can be sold on the market, or stored and used to produce electricity, when the demand on electricity is larger and the prices are higher. The technologies and efficiency of such a process will be discussed.
Further, the economy analysis for hydrogen production and economy analysis for production of electricity from hydrogen in the case of Slovenian electricity market prices will be presented. The economy analysis according to hydrogen technologies in combination with Slovenian nuclear power plant are performed.
_104 ______
AP1000 on Schedule for 2013: Status of Construction in China and USA
Paolo Gaio Westinghouse, Rue Montoyer 10, 1000 Bruxelles, Belgium [email protected]
On July 24 2007 Westinghouse Electric Company signed landmark contracts with China's State Nuclear Power Technology Corporation Ltd (SNPTC), to provide four AP1000 nuclear power plants in China. The four plants are to be constructed in pairs at the Sanmen (Zhejiang) and Haiyang (Shandong) sites. Construction has began in 2009 and the first plant is scheduled becoming operational in late 2013. The remaining three plants are expected to come on line in 2014 and 2015.
The AP1000 is designed to incorporate modern, modular construction techniques. The standard plant is comprised of 50 large and 250 small modules. These modules are constructed in parallel and independent of one another at a shipyard-like factory and later assembled onsite. This technique, together with the simplified plant design, reduces construction costs and schedule. Factory-built modules can be installed at the site with an objective of planned construction schedule of three years - from first concrete pour to fuel load after.
First concrete at Sanmen Unit 1 was completed on March 31st 2009 and, on September 29th 2009, the successful completion of the first pour of basemat structural concrete for the nuclear island at the Haiyang site was announced.
On July 29th 2009, the first of the CA20 nuclear island modules for the two AP1000 reactors under construction at the Sanmen nuclear power plant has been successfully hoisted into place. The module
8 Nuclear Energy f0r New Eur0pe 2010 comprises plant and equipment for used fuel storage, transmission, the heat exchanger and waste collection and is the largest component to be used in the construction of the Sanmen AP1000 units. The containment vessel bottom has been installed afterward and containment erection is proceeding.
The manufacturing of large components for the first two AP1000 units has already started.
The paper presents the status of the AP1000 construction program, including an overview of the activities at both China sites visualized also with pictures and short movies. It will also describe the status of constructions in the USA sites where AP1000 are planned to be built and finally will give an overview prospective of potential construction of AP1000 worldwide.
_105 ______
Development of a Simulation Tool for a Preliminary Analysis of the MSR Core Dynamics
Carlo Fiorina, Antonio Cammi, Claudia Guerrieri, Lelio Luzzi, Benedetto Spinelli Politecnico di Milano – Department of Energy Nuclear Division-CeSNEF Via La Masa, 34 – 20156 Milano, Italy [email protected]
The MSR (Molten Salt Reactor) is one of the six innovative concepts of nuclear reactors envisaged by the GIF (Generation IV International Forum) initiative for the long term evolution of the nuclear technology, in the direction of a more sustainable and economic power generation.
The MSR is characterised by a complex and highly non-linear behaviour, which requires a careful investigation as a consequence of some unusual features like the presence of a fluid fuel and the precursors drift. In this paper, the MSR core dynamics is analysed with reference to the MSRE (Molten Salt Reactor Experiment). Such reactor was a prototype built in the sixties at ORNL (Oak Ridge National Laboratory) and it is frequently adopted as reference thanks to the availability of experimental data.
Numerical models featured by increasing complexity are presented. In particular, zero-dimensional models have been developed according to different approaches for the choice of the state variables. Other models were developed introducing a one-dimensional discretization for the heat convection, the precursor drift or both.
A preliminary comparison between the developed models has been carried out considering also a dedicated model developed at ORNL for the MSRE. Such comparison has shown noteworthy similarities between different models as well as some weakness in the simplest ones.
The variety of developed models represents a starting point in the creation of a complete simulation tool for MSR dynamics, suitable for calculations with different degrees of reliability and numerical complexity.
9 Nuclear Energy f0r New Eur0pe 2010
_106 ______
Appropriate Thermodynamic Cycles to be Used in Future Pressure-Channel Supercrical Water-Cooled Nuclear Power Plants
Laure Lizon-A-Lugrin1, Alberto Teyssedou1 and Igor Pioro2 1 Nuclear Engineering Institute, Engineering Physics Department, École Polytechnique de Montréal, Montréal, Québec. 2 Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Oshawa, Ontario.
The International Atomic Energy Agency (IAEA) has recently stipulated that by the year 2030 world primary-energy requirements will increase by up to 45%. Further, nowadays observed trends in energy supply and consumption do not satisfy environmental sustainability. Therefore, to assure both a healthy world economy as well as adequate social standards, in a relatively short term new energy- conversion technologies are mandatory. To fulfil this requirement, the participation of 10 countries has recently made it possible to establish a Generation IV International Forum (GIF). Thus, GIF members’ have proposed the development of new generation of nuclear-power reactors to replace the present technologies. The principal goals of these nuclear-power reactors, among others are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation.
As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU- type Super Critical Water-cooled Reactor (SCWR), i.e., a pressure-channel or pressure-tube reactor that will use supercritical water as core coolant. Such a system must run at a coolant outlet temperature of about 625oC and at a pressure of 25 MPa. It is important to remark that the use of pressure tubes instead of a pressure vessel will permit the use of advanced steam-reheat thermodynamic cycles. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will largely compete with actual supercritical “steam”-power boilers.
In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be avoided. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. To this end, however, there is still a great amount of work that must be carried out to establish the most reliable and optimal thermodynamic-cycle topology that should be appropriate to future pressure-channel SCWRs.
It must be pointed out that several steam-cycle arrangements used in existing thermal-power plants have been discussed by many authors, but none of the proposed cycles have been appropriately optimized and adapted to the pressure-channel SCWR concept. Thus, the present work is intended to fulfill this gap by including at least two alternative solutions of SCWR power cycles that will use steam reheat in the reactor core to achieve an effective electrical/mechanical power of 1200 MW.
This paper presents corresponding thermodynamic models for preselected cycles as well as their validation among existing data and their optimization using an “evolutionary optimization” technique based on generic algorithms.
10 Nuclear Energy f0r New Eur0pe 2010
_107 ______
Nuclear Infrastructure and Institutional Capacity Restarting a Nuclear Program: The Italian Case
Lorenzo Antola, Andrea Cambriani, Katia Slavcheva, Fabrizio Trenta ITER-Consult, P.zza San Pietro, 36 00029 Vicovaro - Roma [email protected]
After the Chernobyl accident, Italy was the only European country that arrived at the decision to stop the NPPs in operation or under construction and all nuclear projects in progress. After a period of about twenty-five years the Italian Government has reconsidered the use of nuclear energy and has developed initiatives at level of legal-institutional framework and at level of international industrial agreements to launch a new nuclear program.
The Italian Nuclear Infrastructure (NI) was created in the 1960s including the legal and institutional framework and the establishment of an independent regulatory system. With the decision to launch a new nuclear program the NI needs to be revisited, strengthened and made suitable for the new program.
In the last 20 years the Italian major nuclear company Ansaldo Nucleare has kept its activity mainly working on international projects thus remaining active and maintaining knowledge and competencies in the nuclear field.
The Italian major utility, ENEL, after a long absence, has re-entered the nuclear field since few years by creating joint ventures or entering into ownership of utilities outside Italy. The country industrial capability for services, supply of equipment and construction has maintained some capacity and links with the nuclear business, but it needs to carry out a qualification process in order to be competitive with other similar capacities existing in Europe and to achieve a significant participation in the new nuclear projects.
New programs and resources are also going to be planned and made available for research and development (ENEA) and in the field of education (Universities).
The institutional part of the infrastructure, which includes mainly the legal framework and the regulatory system, is undergoing a substantial re-organization process according to the two new laws of 2009 and 2010. In particular the licensing process for new NPP has been revised and made more suitable for today needs. The Nuclear Safety Authority (NSA), according to the July 2009 law, is going to be re-established to face, in more effective way, the new program. The NSA will enter a process of increased regulatory and licensing demand, in addition to the ongoing duties (decommissioning, research reactors, transport of radioactive materials, waste management,…).
The new tasks include the important efforts in establishing safety criteria, develop guidance, establish suitable internal procedures and organization and ensure a transparent communication and information to the public. For that the Italian NSA needs additional resources, development of lost competences, increased management capabilities and updating in the field of safety conception and requirements of current NPP technology and design. It can be foreseen that the NSA will need a substantial support in the nearest future. An important role will be paid by the, appropriate and qualified resources available in the country, together with an effective international cooperation.
ITER-Consult (Ltd), expert organization created in 2003, and ISO certified, has well developed capabilities, and no conflict of interest, to provide independent technical evaluation and support to
11 Nuclear Energy f0r New Eur0pe 2010
NSAs, to maintain safety culture and knowledge, to transfer know how and to actively contribute to Education &Training activities.
Moreover, ITER-Consult, since the very beginning, has recognized and promoted the importance, of international cooperation and networking, establishing relations with international organizations (IAEA, NEA) and European NSAs and TSOs. This is particularly necessary to maintain competence and improve knowledge, to optimize the resources, to get acquainted with different approaches and to benefit from the experience coming from other organizations.
_108 ______
Generation IV NPPS: Supercritical “Steam” Rankine Thermodynamic Cycles Options
Pioro, I., Mokry, S., Miletic*, M., Grande, L., Saltanov, Eu, and Peiman, W. Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology 2000 Simcoe Str. North, Oshawa, Ontario, L1H 7K4 Canada E-mails: [email protected]; and [email protected] * Czech Technical University, Prague, Czech Republic
Research activities are currently underway worldwide to develop Generation IV nuclear reactor concepts with the objective of improving thermal efficiency and increasing economic competitiveness when compared to that of modern thermal power plants. There are six Generation IV nuclear-reactor concepts under development worldwide: 1) Gas-cooled Fast Reactors (GFRs) or High Temperature Reactors (HTRs); 2) Very High-Temperature gas-cooled Reactors (VHTRs); 3) Sodium-cooled Fast Reactors (SFRs); 4) Lead-cooled Fast Reactors (LFRs); 5) Molten Salt-cooled Reactors (MSRs); and 6) SuperCritical Water-cooled Reactors (SCWRs).
The main objectives for developing and utilizing these reactors are: 1) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30 – 35% to approximately 45 – 50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs.
Generation IV NPPs will have much higher operating parameters compared to current NPPs (i.e., outlet temperatures above 500°C). Furthermore, these reactors operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis.
To decrease significantly the development costs of Generation IV NPPs, to increase their reliability, and to achieve similar high thermal efficiencies as the advanced fossil steam cycles, it should be determined whether these NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to that, they can reach thermal steam- cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43% on a Higher Heating Value (HHV) basis).
This paper presents and discusses several Generation IV NPPs layouts and, corresponding to that, thermodynamic-cycle options. An analysis of main parameters and thermal efficiency performance of Generation IV NPP concepts based on a single-reheat regenerative cycles are discussed.
12 Nuclear Energy f0r New Eur0pe 2010
The single-reheat cycles are comprised of: a SCWR, a SC turbine (consisting of one High-Pressure (HP) cylinder, one Intermediate-Pressure (IP) cylinder and two Low-Pressure (LP) cylinders), one deaerator, several feedwater heaters, and pumps. Since the single-reheat option includes a “nuclear” steam-reheat stage, the SCWR is based on a pressure-tube design. A thermal-performance simulation reveals that the overall thermal efficiency is approximately above 50%.
Current studies have shown that indirect cycles with single-reheat configurations might be the best choice or at least interim choice for the Generatioj IV NPPs concepts. However, the indirect single- reheat cycle requires large heat exchangers, thus increasing the complexity of the NPP design.
_110 ______
Estonia: The Prospects for Nuclear Energy
Raphael J. Heffron Judge Business School University of Cambridge, Cambridge, CB2 1AG, United Kingdom [email protected]
Kalev Kallemets Department of Economics Tallinn Technical University, Tallinn, Estonia [email protected]
This research examines energy policy in the EU with the focus being on nuclear energy. It assesses the nuclear energy option in comparison to other energy choices. With the change that has occurred worldwide in terms of climate change, nuclear energy has become an important energy option for any country in the EU even the smaller countries. This paper assesses nuclear energy as an option for small EU countries with Estonia being the country researched in depth.
The research assesses the changing conditions for new nuclear build in European Union (EU) from a large, medium and small EU country perspective. This paper focuses on the latter category and examines what has occurred in Estonia where nuclear energy presents itself as the leading energy source in the twenty-first century. The paper examines Estonian economic development and the political and historical issues that have created a successful Eastern European EU new entrant. The research revolves around interviews with key stakeholders. It continues with as assessment of how nuclear energy represents the best option for energy in Estonia from a myriad of perspectives.
The results attest that due to various political, historical, strategic, social and economic reasons nuclear energy presents itself as an exciting option for Estonia. This case of Estonia provides a lesson for other EU countries that have no nuclear power but where it should at least be considered in their respective energy mixes. As the argument for nuclear energy grows stronger in Estonia, this research also states some of the exciting opportunities that will result for the Estonian economic sector.
13 Nuclear Energy f0r New Eur0pe 2010
_111 ______
Investigations in the Near and Long Term Perspective of Italian Scenario
R. Calabrese ENEA Via Martiri Di Monte Sole 4, 40129 Bologna, Italy [email protected]
The change in the Italian nuclear policy aims at achieving 25% nuclear electric capacity share by 2030. According to an annual electricity demand increase of 1%, in a low scenario, and 2.5%, in a high scenario, the estimated installed capacity will roughly span from 12 GW(e) to 20 GW(e). To accomplish this target different combinations of thermal reactors technologies are possible. In this paper it is intended to investigate the impact of different mix in LWR fleet on aspects such as natural uranium consumption, spent fuel amount, plutonium and minor actinides (MA) inventories, radiotoxicity.
In the hypothesis of nuclear fuel cycle closure, aiming at recycling plutonium and MA, the Italian scenario is also investigated in the long term to estimate the deployment of fast nuclear systems with different breeding ratios. On this regard, besides previous parameters, great attention is paid to the needed reprocessing plant capacity, the fast systems deployment date and, finally, the effect of thermal systems introduction rate on the planning of fuel cycle closure.
The analysis is performed by means of the DESAE code, a tool developed within the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) of IAEA.
14 Nuclear Energy f0r New Eur0pe 2010
Reactor Physics, Fuel Cycle and Research Reactors
15 Nuclear Energy f0r New Eur0pe 2010
_201 ______
Spectral Codes Pin Power Prediction Comparison
Davor Grgić, Radomir Ječmenica, Dubravko Pevec Faculty of Electrical Engineering and Computing, University of Zagreb Unska 3, 10000 Zagreb, Croatia [email protected], [email protected], [email protected]
Nodal neutron diffusion codes relay, in their pin powers predictions, on data pre-calculated on fuel assembly basis using 2D transport codes. That is true for both, steady state (design and in-core fuel management) and transient, core pin power distribution. Beside the assumptions used during pin power reconstruction, final results depend on the accuracy of performed 2D transport depletion calculation. In the paper different 2D spectral codes were used to calculate pin power distribution for reflected single fuel assemblies during depletion. The assemblies are NPP Krsko 16x16 fuel assemblies without and with IFBAs, for two U-235 enrichments 4.4% and 4.8%, as found in cycle 24 of the plant. The heterogeneous depletion calculation was performed up to burnup of 60 GWd/tU. Four spectral codes were used:
DRAGON (Version4) is neutron transport (collision probabilities and method of characteristics) code for calculation of pin cells and fuel assemblies developed at Institut de Genie Nucleaire Ecole Polytechnique de Montreal,
FA2D is 2D transport collision probability code developed at Faculty of Electrical Engineering and Computing, University Zagreb,
NEWT (as found in SCALE 6.0) is discrete ordinates transport code developed at Oak Ridge National Laboratory, and
PSG2/Serpent is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, specialized in two-dimensional lattice physics calculations, developed at VTT Technical Research Centre of Finland.
The prediction capabilities of the codes are compared for infinite multiplication factor, and for 2D pin power form factors.
_202 ______
Building a Dynamic Monte Carlo Code to Simulate New Reactor Concepts
N. Catsaros, B.Gaveau, M.Jaekel, J. Maillard, G.Maurel, P.Savva, J.Silva, M.Varvayanni,
During the last decades, innovative and –to a certain extent- “exotic” nuclear reactor designs have been proposed, addressing particular problems such as the management of nuclear wastes. These designs include the Accelerator Driven Systems (ADS), the “candle” reactors etc.
These innovative reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. In fact, the ability of a modern code to take into account the system’s evolution, is mandatory. A continuous feedback procedure must be established between the main inter-related parameters of the system such
16 Nuclear Energy f0r New Eur0pe 2010 as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADS are concerned, the ability of the code to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required.
Making profit of the modern processors’ enhanced capabilities, a new Monte Carlo code is under development, based on the GEANT3 High Energy Physics code and satisfying all the above requirements. A rough description of the implemented methodologies is given in this work, together with some illustrative applications of the code.
_203 ______
Neutronics Analysis of the TRIGA Vienna Mixed Core
H. Böck, R. Khan Vienna University of Technology/Atominstitute Stadionallee 2, A-1020, Vienna, Austria [email protected], [email protected]
The 250kW TRIGA Mark II research reactor, Vienna, operates since March 7th 1962. Originally it was started up with type 102 aluminium clad fuel. Later on due to fuel consumption SST clad type 104 fuel elements have been added to compensate for burn-up. Finally in 1975 HEU TRIGA fuel (FLIP fuel = Fuel Lifetime Improvement Program) was available from General Atomics:Nine FLIP fuel elements were purchased to assure long-term reactor operation without refuelling. Therefore today the TRIGA reactor Vienna operates therefore with a completely mixed core using these three types of fuel elements with two different enrichment. This makes burn-up and core calculations very difficult. The present work presents the results of these core calculations using MCNP5 modelling. To validate the calculations several spent fuel elements including one FLIP fuel have been scanned axially using a unique scanning device developed at the Atominstitute.
_204 ______
Methodology Investigations on Uncertainties Propagation in Nuclear Data Evaluation
P. Dossantos-Uzarralde, H. P. Jacquet, G. Gauriot
Nuclear data evaluations refer to a body of techniques for investigating the phenomena of propagation of parameter uncertainties into nuclear reaction models. A classic method consists in testing the technique on the collection of data through observation and experimentation. However procedure hypotheses vary from one technique to another. A basic expectation required by users of the results is the opportunity to estimate the quality of the information, while having the conviction that the technique is objective to reduce biased interpretations of the results.
In this paper, we discuss distinct ways of calculating the nuclear cross section variance-covariance matrices issued from the spherical optical model : least square method, Bayesian method but also the spectral methods using a polynomial decomposition. We attempt to evaluate the differences between the uncertainties generated by these methods. Discussing the results of simulation for the calculation of all major neutron reactions on 89Y and 196Au, our interest was focused on the probability distribution function used in the Monte Carlo technique for quantifying covariance matrices. In particular we
17 Nuclear Energy f0r New Eur0pe 2010 discuss how parameter multivariate probability function distribution is applied to the problem of propagating errors. We also discuss the use of statistical regressions of the randomly selected values of the uncertain parameters to determine the importance of parameters contributing to the final uncertainty.
_205 ______
DRAGON and CORD-2 Nuclear Calculation of the NPP Krško Fuel Assembly
Marjan Kromar “Jožef Stefan” Institute Reactor Physics Division Jamova 39, 1001 Ljubljana, Slovenia [email protected]
Bojan Kurinčič Nuclear Power Plant Krško Engineering Division - Nuclear Fuel & Reactor Core Vrbina 12, 8270 Krško, Slovenia [email protected]
The geometry of the reactor core is too complex to be solved in one step. Therefore, a solution for the whole core in 3-D geometry is sought in several steps, where some kind of homogenization procedure of neutron few-group cross sections is applied. Usually, assembly-homogenized effective two-group cross sections are determined, which are suitable for solving the diffusion equation for the whole core by a coarse mesh nodal methods. In this paper DRAGON 4 and CORD-2 codes are used for the calculation of NPP Krško 16×16 fuel assemblies without and with IFBAs. DRAGON code was selected, since it can use the same cross-section library as the WIMS-D5 code used in the CORD-2 system. Different results arise therefore solely from the different models used in the calculations. The heterogeneous depletion calculation was performed up to burnup of 60 000 MWd/tU. Results of both codes are compared for infinite multiplication factor and selected 2-group homogenized cross section constants.
_206 ______
Evaluation of Integral Benchmark Experiments Uncertainty due to Boron Isotopic Abundance Variations
Gašper Žerovnik, Luka Snoj “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
The isotope 10B is a strong thermal neutron absorber, which along with its relative inexpensiveness and availability makes boron a very important material in nuclear engineering. For most applications, natural boron is used, though use of 10B enriched (the remainder is 11B) is also applicable. Thermal 10B absorption cross section is five orders of magnitude larger than the corresponding 11B cross section, therefore macroscopic boron (mixture) cross section and consequently integral reactor parameters, 10 11 such as keff, are very sensitive to the B/ B ratio.
18 Nuclear Energy f0r New Eur0pe 2010
Natural abundance of the isotope 10B in boron ranges from 18.9 % to 20.3 % and strongly depends on the location of the boron deposit (mine). Since this interval can be labeled as “100 % confidence interval”, the realistic boron sample (standard) uncertainty is much smaller. With a detailed study [1] of boron isotopic composition regarding its origin, the 10B abundance distribution was found to be asymmetric and strongly deviating from normal distribution. Furthermore, the uncertainty in isotopic abundance has been reduced to 0.09 % with a mean value of 10B abundance of 19.92 %. If the exact origin of the boron is known, this uncertainty can be further reduced.
Uncertainty in boron concentration and isotope composition contributes a significant part to the total experimental uncertainty of many integral reactor experiments. E.g., in the most comprehensive compilations of criticality and reactor physics experiments, ICSBEP [2] and IRPhEP [3] handbooks, there is a number of experiments with experimental uncertainties resulting mostly (or at least significantly) from uncertainty in the isotopic composition of the boron used in the experiment. Current approach, using the [18.9,20.3] % interval for 10B abundance as the bounding (3σ) uncertainty, overestimates the uncertainty due to B isotopic composition. Within handbooks [2] and [3], examples have been found where using the reduced uncertainty 10B content, total uncertainties of integral parameters can be reduced.
[1] A. Lukek, L. Snoj, G. Žerovnik, Natural variations in boron isotopic abundance, IJS work report 10275, Ljubljana, Slovenia (2009).
[2] International Handbook of Evaluated Critical Safety Benchmark Experiments, Organization for Economic Cooperation and Development - Nuclear Energy Agency, NEA/NSC/DOC(95)03, Paris, published on DVD, ISBN 978-92-64-99054-8 (2009). [3] International Handbook of Evaluated Reactor Physics Benchmark Experiments, Organization for Economic Cooperation and Development - Nuclear Energy Agency, NEA/NSC/DOC(2006)1, Paris, published on DVD, ISBN 978-92-64-99055-5 (2009).
_207 ______
Modeling of Pool Critical Assembly Pressure Vessel Facility Benchmark
Dinka Vragolov University of Dubrovnik Department of Electrical Engineering and Computing Ćira Carića 4, 20000 Dubrovnik, Croatia [email protected]
Mario Matijević, Dubravko Pevec Faculty of Electrical Engineering and Computing Department of Applied Physics Unska 3, 10000 Zagreb, Croatia [email protected]; [email protected]
A Pool Critical Assembly Pressure Vessel Facility Benchmark [1] has been chosen for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the U.S. Nuclear Regulatory Commission regulatory guide [2].
The SCALE (Standardized Computer Analysis for Licensing Evaluation) [3] code package, developed at Oak Ridge National Laboratory, was used for modeling of the Pool Critical Assembly Pressure Vessel Facility Benchmark. The CSAS6 sequence of the SCALE code package, which includes
19 Nuclear Energy f0r New Eur0pe 2010
KENO-VI Monte Carlo code, was used for calculation of equivalent fission fluxes. The calculations were performed using different multi-group cross section libraries.
The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.
[1] I. Remec and F.B. Kam, “Pool Critical Assembly Pressure Vessel Facility Benchmark”, NUREG/CR-6454 (Prepared for NRC by Oak Ridge National Laboratory, ORNL/TM-13205), July 1997.
[2] “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence” , Regulatory Guide 1.190, U.S. Nuclear Regulatory Commission, March 2001.
[3] “SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation“, ORNL/TM-2005/39, Radiation Safety Information Computational Center at Oak Ridge National Laboratory.
_208 ______
Linac Based Subcritical Th232-U233 Research Reactor
Ali Pazirandeh, Elham Younesian Science and Research Branch, Islamic Azad University Tehran Iran [email protected]; [email protected]
With growing demand for scientific and technical training of nuclear engineers for Nuclear Power Plants in Iran, there is a great need for design and construction a research reactor. With the view of Non-Proliferation Treaty, Th-U233 fuel cycle is quite justifiable.
High-energy electrons from a compact linac impinge on a tungsten target produce high-energy bremsstrahlung radiations. The radiations interaction with lead and beryllium produce high-energy neutrons, which are slowed down in water as moderator and coolant. The results of our simulation 8 2 using FLUKA code showed that the system creates thermal neutron flux of 10 n / cm .s after pass through water with a neutron energy spectrum extending in 10-3 eV–5MeV range, with a sharp peak at 0.5eV in the thermal neutron range and very low flux in fast region.
The proposed reactor core is composed of a central tungsten cylinder, 2R=H=10cm as a target for bremsstrahlung radiation, surrounded by fuel rods in 4 annular in a core of radius about 35cm. The total number of fuel rods is 60, which are arranged in four annuli. The height of the core is 80cm. The fuel is Th-U233 oxide with 3% U233. The cladding is made of aluminium 0.6mm thick. The average flux for 200kW reactor power is estimated at 2*1012n/cm2.s
The neutronic parameters calculated using WIMSD4 code. Then feeding two group cross sections in CITATION-LIN2 code two group fluxes were obtained. At steady state core, having considered Xe- 135 and Sm-149 accumulation, the core remains at subcritical throughout its life. By operating reactor at k=0.97, it remains quite safe. However, for safety reason a bank of control rods sustains the core sub-criticality.
20 Nuclear Energy f0r New Eur0pe 2010
_209 ______
MOX Benchmark Calculation Results by Monte Carlo and Deterministic Codes
M. Pecchia, G.Kotev, C. Parisi, F. D’Auria San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa via Diotisalvi 2, 56122 Pisa [email protected], [email protected], [email protected], [email protected]
A depletion calculation benchmark devoted to MOX fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOX fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations.
In the present work the burnup code MONTEBURNS2.0 was used to calculate the masses of inventory isotopes, such code couples the Monte Carlo code MCNP5 to the depletion code ORIGEN2.2 by iterative steps procedure.
To consider the cumulative effect of fission products a new methodology was developed to create a special continuous energy Xsec for lumped fission products.
A comparison of results with deterministic code NEWT/SCALE6 is also presented.
_210 ______
Extension of the MATSSF Code for Self-shielding Factor Calculations of Heterogeneous Samples
Gašper Žerovnik, Andrej Trkov “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
Christophe Destouche Commisariat a l’Energie atomique Centre de Cadarache, DER/SPEx/LDCI 13108 Saint Paul les Durance, France [email protected]
A neutron activation experiment was performed in a research reactor at CEA, Cadarache by irradiating 11 sample foils, made of different materials, at the same time. Within the same irradiation channel, the foils were positioned in two separate stacks contained in aluminium capsules, including 5 and 6 samples, respectively. Since the sample foils are relatively thick (their mean chord length is less than an order of magnitude smaller compared to the neutron attenuation length) and in direct contact within the capsule, separate sample treatment underestimates the self-shielding effect. Program MATSSF (avalable at: http://www-nds.iaea.org/naa/matssf/), which originally calculates self-shielding factors for cylindrical homogeneous samples (foils or wires), has been extended to heterogeneous geometry (i.e. to a finite number of homogeneous cylindrical samples) and verified on the realistic experimental configuration as described above. The method will be presented in the full paper. As a reference,
21 Nuclear Energy f0r New Eur0pe 2010
Monte Carlo calculation with MCNP5 code has been taken, using a full geometry model of the irradiation channel including all details within the channel, and a neutron source with multigroup spectrum, specific to this irradiation facility.
22 Nuclear Energy f0r New Eur0pe 2010
Reactor Operation
23 Nuclear Energy f0r New Eur0pe 2010
_301 ______
Operational Know-How and Know-Why – AREVA´s Internal and External Exchange of Information and Experience
Dipl.-Ing. Peter SCHIMANN AREVA NP GmbH Installed Base Engineering IBE1-G Koldestraße 16, D-91052 Erlangen, Germany [email protected]
Zoran V. STOSIC AREVA NP GmbH Director R&D Installed Base Engineering Director Marketing & Sales Ex-Yugoslavia Koldestraße 16, D-91052 Erlangen, Germany [email protected]
In the past decades the nuclear option has developed to become a measure to secure the energy supply not only of the established nuclear nations, but also of newly embarking or smaller countries without significant nuclear experience. In these small countries cooperation may work cross-border between countries or cross-corporate between manufacturers and/or utilities.
Experiences gathered in all phases of a nuclear project have to be applied on all levels to support the construction of absolutely safe and safe operation of nuclear reactors. Since the time span between decision-making for building, the licensing procedure, the detailed planning, the construction phase and commissioning can amount to several years, this leads to the urgent requirement of having to incorporate the constant experience gain on all technical and administrative levels. This has to be taken into account for the safety-related boundary conditions in the frame of a new build project as well as for any kind of updating measures for existing plants. Often neglected is the fact that this also applies to the choice of the guidelines to put into effect, to an efficient and competent authority and utility body and the integration of adequate technical specialists. Utilities face different requirements depending on whether they are situated in the country of responsible nuclear power plant utilities’ HQs or whether they are merely co-owners of a foreign plant. All these matters profit from resorting to the knowledge pool from operational experiences and experiences gained in new build projects.
AREVA serves the whole nuclear cycling including plant construction and supports its partners in all project phases, even in the development of national institutions and knowledge transfer. Any AREVA knowledge, as recent it may be, is introduced directly into corresponding application fields.
Special attention is drawn to the matter of the resulting operational experience feedback derived from events in nuclear power plants all over the world. The key is learning from mistakes made by others. By quickly adapting improvements can be implemented early in the planning stage of new build projects while at the same time defects and malfunctions in operating plants can be avoided effectively. This means the originally AREVA-internal experience feedback grew to become a knowledge pool available to all our customers. Especially smaller or embarking countries take advantage of this offer to a greater extent since they still lack knowledge. AREVA performs the operational experience feedback for its own product portfolio, in the frame of know-how and know- why transfers and long-term stipulations with customers.
24 Nuclear Energy f0r New Eur0pe 2010
The focus of the following discussion lays on the presentation of the AREVA-internal experience gain and its benefit for our customers in daily business, e.g. by stipulations.
_302 ______
Latest Extensions of the FAR - Software Package for the Nuclear Material Accounting
S. Slavič, M. Kromar, B. Žefran “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
Software package FAR (Fuel Assembly Register), developed by the Reactor Physics Department of the Jožef Stefan Institute, covers all aspects of the nuclear material accounting on a nuclear facility. The main purpose of the package is to simplify control and update data base of the fuel elements and their properties during NPP lifetime. It enables automatic creation of various reports in specific template forms suitable either for internal NPP use or forms that satisfy strict international reporting requirements. Program package runs on PC machines under Windows operating system. The code is under permanent development to match all customer needs in reasonable time.
Latest extension of the program takes into account also individual fuel rods data. Fuel rods extracted from fuel assemblies are stored in the special fuel rod storage basket (FRSB). FRSB is suitable also for storage of stainless steel rods, that are used as replacement rods in the fuel assemblies' reconstitution process.
Characteristic data of individual fuel rods and stainless steel rods are kept in additional data base file (DBF). For each rod the following items are stored: • name of the root fuel assembly, • location of the rod in the fuel assembly, • location of the rod in the basket, • nuclear data of the fuel rod (burnup, U-total, U235, enrichment, Pu%, Pu-total).
Layout of the rods in the basket can be presented in a graphical print form or displayed directly on the monitor.
_303 ______
® Ovation to Third Party Interface Solutions
Drazen KrajinaPWR DCIS Upgrades - DC&IS Europe/South Africa [email protected]
Dave Bubeck Repair Replacement & Automation [email protected]
Westinghouse Electric Company LLC, headquartered in Cranberry, Pennsylvania, USA with the focus of three core businesses of Westinghouse — Nuclear Fuel, Nuclear Services and Nuclear Power Plants, supply leading-edge nuclear technology to satisfy the world's ever growing global demand for clean, safe and reliable nuclear energy. Westinghouse technology is the basis for over 40% of the
25 Nuclear Energy f0r New Eur0pe 2010 world's operating nuclear plants. Nuclear Services, specifically Instrumentation & Controls (I&C), is one area in which Westinghouse is a world leader, as well as a competitor in the U.S. and global market. Currently, one of the main drivers for solutions in the I&C market is created by the outdated technology in aged nuclear power plants. With the proper I&C upgrade technology; existing power plants can enhance their operations and increase production of nuclear energy.
® Westinghouse has established a competitive technology using the Ovation product to interface with third-party systems and has developed new concepts and solutions in the shifting I&C market. Using ® Ovation Controller Modules Westinghouse developed their ability to facilitate the execution of ® interfaces between a third-party system and Ovation with products such as, the Link Controller, Ethernet Link Controller and Smart Instrumentation Buses. Furthermore, Smart Instrumentation Buses such as HART I/O, DeviceNet, ProfibusDP and Foundation Fieldbus H1 are connectivity options supported by Westinghouse for a third-party interface. These communication systems not only serve as a base level network, but they enable the development of custom interfaces to third-party systems while sustaining redundancy at the module level.
While remaining competitive with industry standard connectivity options, Westinghouse uses several ® protocols to enable data transfer between Ovation and a third-party system. Specifically, the Data Link Server, developed by Westinghouse, supports Modbus Master, WESGate client, FTP/SFTP (via formatted text files), BEACON, Allen Bradley, APP CTS/CTA, GE Mark V&VI Trubine Control, MHI Turbine Control, DRCS (Westinghouse proprietary interface) and AIO Gateway (Westinghouse proprietary protocol). In addition, Westinghouse uses several custom protocols that vary based on ® customer needs. Namely, the Data Link Server permits custom interfaces between Ovation and a third-party system.
The Data Link Server is built on a server platform enabling the combined installation of Application Server and Data Link Server. Utilizing the Application server features by monitoring the status of the ® Data Link Server on the Ovation Highway, monitoring of executed applications, management of applications and logging of exceptions and events, enables the Data Link base software to facilitate custom Data Link development for a customer's necessity.
The flexibility of these products allows plants to complete a phased-upgrade program, rather than committing to the large investments required for complete control system upgrades.
_304 ______
Investigation of Grid-To-Rod-Fretting at Krško Nuclear Power Plant
M. Chambers, D. Kadivnik, B. Kurinčič Nuklearna Elektrarna Krško, Vrbina 12, 8270 Slovenia [email protected], [email protected]
Ever increasing values power output and cycle length (i.e. burn up) are placing more and more demands on Fuel Assembly (FA) design limits and its integrity as a barrier. FA design is currently not proactive and is problem led, resulting in long lead times design changes to ever more rapidly increasing demands. Over the last 30 years, fuel failure causes have changed dramatically. Grid-to- rod-fretting currently represents the largest cause of fuel failures (greater than 70% in 20071) for light water reactors and has been for the last 10 years. There are many possible causes for GTRF issues where all can be related to a physical fuel assembly design failure to non-envisioned operational
26 Nuclear Energy f0r New Eur0pe 2010 circumstances. The ambitious “INPO 0 by 2010” statement2 has not been met by a large amount of utilities.
Krško nuclear power plant (Krško NPP) after power uprate (steam generator replacement) had 3 non- leaking cycles (17-19) before fuel failures were observed in chemical coolant activity in cycle 20. The succeeding 3 cycles (cycles 20-23) were all leaking. Post irradiation examinations routinely performed after core unloads and reconstitution examinations after cycle 22 found the cause to be GTRF. This contribution details the analysis and investigation by Krško NPP in the prevention of the GTRF mechanism of failed FA.
References
1. The Path to Zero Defects: EPRI Fuel Reliability Guidelines, Electric power research institute (EPRI) executive summary, (march 31, 2008).
2. Institute of Nuclear Power Operations (INPO) statement, 2006.
3. NEK Post irradiation examination inspection results.
4. Root cause analysis for leaking fuel assemblies in NPP Krško cycles 20-23.
5. NEK Chemical coolant analysis, cycles 20-23.
_305 ______
New Control and Data Acquisition of Experiments at VR-1 Training Reactor
Martin Kropik, Jan Rataj Department of Nuclear Reactors Faculty of Nuclear Sciences and Physical Engineering Czech Technical University in Prague V Holesovickach 2, CZ 180 00 Prague 8, Czech Republic [email protected], [email protected]
The paper deals with the new control and data acquisition of experiments at VR-1 training reactor in Prague. The control and data acquisition of experiments was significantly improved by a new human-machine interface of the reactor, a new History server and a new computer for evaluation of experiments.
The new human machine interface provides better control of the reactor operation than the previous one. The graphical presentation of the reactor status was improved significantly, new commands to facilitate the reactor running were implemented, and the English user interface important for foreign users was enhanced. Also the batch commands operation of the reactor (edit and run of batches) was notably extended.
The new History server connected with the PC of the human-machine interface stores all operational data of the reactor every 0.1 second (the standard I&C cycle). These data could be presented either at the human machine interface or can be sent to the new computer for evaluation of experiments. The storage capacity of the History server makes possible to store operational data at least for ten years. Parallel hard discs (Raid 1) are used for a safe data storage. The operational data can be used either for evaluation of experiments or for the reactor I&C calibration or for the documentation of the safe reactor operation.
27 Nuclear Energy f0r New Eur0pe 2010
The new computer for evaluation of experiments is connected to the human-machine interface. In order to provide the human-machine interface and all reactor I&C security, the communication is based on a one way fibre optics line with output from the human-machine interface. This measure provides protection of the human-machine interface and the I&C against external harmful influences, e.g. viruses or spyware. The high speed serial card MOXA with fibre optics interface and baud rate up to 921 kb/s was used for communication. The new software Experimental Studio from dataPartner company has been installed on this computer. The Experimental Studio software very significantly supports the evaluation of reactor experiments. It is possible to prepare tasks as a combination of standard programming languages (C#, Visual Basic Jscript, IronPython, etc.) and graphical oriented tools.
The above mentioned facilities with their software significantly enhance the control of the VR-1 training reactor operation, provide an improved documentation of the safe reactor operation, facilitate the maintenance and calibration of the reactor I&C, offer better data acquisition during reactor experiments and also the evaluation of them.
28 Nuclear Energy f0r New Eur0pe 2010
Thermal Hydraulics
29 Nuclear Energy f0r New Eur0pe 2010
_401 ______
The Impact on Flow Meter Uncertainty of Tubular Flow Conditioners
Ernest M. Hauser [email protected] Cameron Coraopolis, PA, USA
Herbert Estrada [email protected] Cameron
Flow meters of various types have long been known t Coraopolis, PA, USAo be sensitive to upstream hydraulic configurations which can adversely affect fluid velocity profiles.
Differential pressure devices can be affected by both swirl and non-symmetric velocity profiles. Turbine meters can be biased by non-axial velocity attack angles on the rotating blades, and single and multi-path ultrasonic meters have varying sensitivities to fluid velocity profiles de[ending on the specifics of their configuration.
Tube type and plate type flow conditioners are often used to create a uniform, repeatable and predictable velocity profile to eliminate biases in these measurement devices. For example, nozzles designed to meet the requirements of ASME PTC-6 are required to be equipped with upstream flow tubes with tube type flow conditioners.
Conventional wisdom holds that tube and plate type flow straighteners condition and control the fluid velocity profiles downstream so that, when they are installed, the downstream flow instrument, when calibrated with the straighteners, will be less sensitive to upstream hydraulics, and therefore deliver a reduced measurement uncertainty.
The paper analyzes detailed velocity profile data from laboratory testing and field installations which challenge these assumptions. It evaluates the uncertainties introduced by variations in hydraulic configuration itself, both of the conditioner proper and of the upstream fluid system. And it analyzes time-wise variations velocity profiles brought about by the flow conditioners themselves, apparently due to corrosion product deposition similar to that which occurs in the throats of flow nozzles and orifice plates.
30 Nuclear Energy f0r New Eur0pe 2010
_402 ______
Steam Bubble Condensation in Polydispersed Flow – Experiments and CFD Simulations
Eckhard Krepper, Martin Schmidtke, Dirk Lucas, Matthias Beyer Forschungsuentrum Dreden-Rossendorf e.V. P.O.B. 510119, D-01314 Dresden, Germany [email protected]
Bubble condensation in sub-cooled water is a complex process, to which various phenomena contribute. Since the condensation rate depends on the interfacial area density, bubble size distribution changes caused by breakup and coalescence play a crucial role.
To investigate the involved phenomena and their complex interplay, experiments on steam bubble condensation in vertical co-current steam/water flows have been carried out at the TOPFLOW test facility in FZD, which consists of an 8m long vertical DN200 pipe. Steam is injected into the pipe and the development of the bubbly flow is measured at different distances to the injection using a wire mesh sensor. By varying the steam nozzle diameter (1mm or 4 mm) the initial bubble size can be influenced. Larger bubbles come along with a lower interfacial area density and therefore condensate slower (see figure). In the experiment besides the vapour volume fraction also the steam velocity, the pressure drop along the pipe as well as the temperature at selected points is measured (Lucas et al., 2009). The additional sensors allow for choosing a defined gas inflow pressure as well as a distinct sub-cooling temperature at the injection location. Steam pressures between 1-2 MPa and sub-cooling temperatures from 2 to 4 K were applied. Due to the drop of hydrostatic pressure along the pipe, the saturation temperature falls towards the upper pipe end. This affects the sub-cooling temperature and can even cause re-evaporation in the upper part of the test section.
Figure: Evolution of the cross-section averaged gas volume fraction along the pipe for two different steam injection orifice sizes In the present contribution, the condensation experiments are compared to simulations using an extended MUSIG approach in the CFD code CFX-12.1. This approach has been developed in cooperation with ANSYS-CFX for the computation of condensation in polydispersed bubbly flows with CFD. This extended MUSIG approach includes the transition of bubbles to smaller size groups due to condensation as well as the shift of the bubble size distribution due to pressure changes. The new CFD approach is able to catch all relevant phenomena at least qualitatively, such as bubble condensation and evaporation and radial bubble size segregation. Crucial for the condensation simulations is the modelling of coalescence and breakup, which still needs to be improved. 31 Nuclear Energy f0r New Eur0pe 2010
References
Lucas D. and Prasser H.-M., Nuclear Engineering and Design, Vol. 237, Issue 5, pp. 497-508, 2007.
Lucas D., Beyer M., Frank T., Burns A. and Zwart P., Article-No. N13P1097, The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) , Kanazawa City, Ishikawa Prefecture, Japan, September 27-October 2, 2009.
_403 ______
Natural Circulation Simulation with Lumped-Parameter Codes Using Input Models Based on CFD Simulation
Ivo Kljenak, Borut Mavko “Jožef Stefan” Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
Simulations of atmosphere mixing and stratification in nuclear power plant containments with a local instantaneous description, using Computational Fluid Dynamics (CFD) codes, have lately become possible with the increase of computer power. However, calculations with these codes still take relatively long times. An alternative faster approach is to model the nonhomogeneous atmosphere with lumped-parameter codes by dividing larger control volumes into smaller ones, in which conditions are modelled as homogeneous. The flow between smaller volumes is modelled using one-dimensional approaches, which usually includes the prescription of flow-loss coefficients. To contribute to the further development of this approach, a modelling of nonhomogeneous atmosphere with lumped- parameter codes is proposed, where the subdivision of a large volume into smaller ones is based on results of CFD simulations. The basic idea is to use results of a CFD simulation to define regions, in which the flow velocities have roughly the same direction. These regions are then modelled as control volumes in a lumped-parameter model. In the present work, the approach is proposed with the lumped-parameter codes ASTEC and CONTAIN. The procedure was applied to a simulation of an experiment on atmosphere mixing and stratification, which was performed in the TOSQAN facility. The facility is located at the Institut de Radioprotection et de Sureté Nucléaire (IRSN) in Saclay (France) and consists of a cylindrical vessel (volume: 7 m3), in which gases are injected upwards at the vessel axis. In the experiment, air was initially present in the vessel, and steam, air and helium were injected during different phases of the experiment. The thermal-hydraulic behaviour was determined by the following dominant physical phenomena: gas injection, steam condensation on some vessel walls, heat transfer and buoyant flow. During certain phases of the experiment, intermediate steady states were reached when the steam condensation rate became equal to the steam injection rate, with boundary conditions remaining constant. In the proposed work, two steady states during which natural circulation was the dominant phenomenon were simulated independently by applying the approach described above, with the code CFX4 used as a CFD code. The simulation with the CFX4 code was performed with a two-dimensional axisymmetric model. Simulations with the ASTEC and CONTAIN codes then succeeded in replicating the basic pattern of natural circulation in the atmosphere of the TOSQAN vessel.
32 Nuclear Energy f0r New Eur0pe 2010
_404 ______
CFD Analysis of Free Surface Flows Using an Open Source Code
Alexander Churbanov Nuclear Safety Institute, Russian Academy of Sciences 52 B. Tul’skaya, 115191 Moscow, Russia [email protected]
Open source CFD codes become more and more popular for numerical studies in science and industry including nuclear reactor safety applications. They provide wide and powerful possibilities to construct and tune specific CFD models, for instance, for modeling of immiscible multi-fluid systems. Such free surface flows occur during a Pressurized Thermal Shock (PTS) situation and require efficient CFD tools for their numerical analysis.
Open source code OpenFOAM [1] designed for computational continuum mechanics has been used successfully in the present work to simulate some free surface flows. Moreover, a cross-verification with commercial CFD code FLUENT [2] has been performed, too. Two validation cases have been studied numerically here on the basis of VOF method.
First, a water jet impinging on an inclined flat plate in air environment has been predicted and compared with measurements. This validation case VAL02 has been proposed in the ECORA Project [3,4] for PTS problems and is based on experimental-numerical work [5]. The 3D steady-state turbulent flow has been predicted with various turbulence models using both codes and compared with measured pressure distribution along the plate.
Next, an unsteady problem of a dam break with a small obstacle placed in the way of the wave front has been calculated. This problem is used as a tutorial in OpenFOAM [6] where photographs from [7] are used for a qualitative comparison only. In the present work more comprehensive experiments performed for this problem in MARIN Institute [8] were employed for validation. Namely, time- histories of the pressure at points of measuring were used for the comparison with 3D transient predictions.
Grid convergence and the influence of the front reconstruction techniques have been studied for both cases. It was found that OpenFOAM provides a good enough agreement with experimental data as well as with predictions via FLUENT and can be utilized to construct more complicated CFD models to predict multiphase flows.
1. OpenFOAM: The Open Source CFD Toolbox. User Guide v.1.5. OpenCFD Ltd. (2008).
2. FLUENT 6.3 User’s Guide. Fluent Inc. (2006).
3. M. Scheuerer. Selection of PTS-relevant test cases. EVOL-ECORA-D05a (2003).
4. M. Scheuerer, M. Heitsch, F. Menter et al. Evaluation of Computational Fluid Dynamic Methods for Reactor Safety Analysis (ECORA). Nucl. Eng. and Design 235 (2005) 359-368.
5. S. Kvicinsky. Methode d’Analyse des Ecoulements 3d a Surface Libre: Application aux Turbines Pelto. PhD Thesis, EPFL, Lausanne (2002).
6. O. Ubbink. Numerical Prediction of Two Fluid Systems with Sharp Interfaces. PhD Thesis, Imperial College of Science, Technology&Medicine, University of London (1997).
33 Nuclear Energy f0r New Eur0pe 2010
7. O. Koshizuka, H. Tamako, Y. Oka. A particle method for incompressible flow with fluid fragmentation. Comput. Fluid Dyn. J. 4 (1995) 29-46. 8. K.M.T. Kleefsman, G. Fekken, A.E.P. Veldman et al. A Volume-of-Fluid based simulation method for wave impact problems. J. Comput. Phys. 206 (2005) 363-393.
_405 ______
Experimental Investigation of Post-Dryout Heat Transfer in Annulus With Flow Obstacles
I.Anghel, H. Anglart, S.Hedberg KTH-Albanova, Roslagstullbacken 21, Stockholm 10691, Sweden [email protected], [email protected], [email protected]
The purpose of this paper is to present the results of the post dryout heat transfer experiments performed in an annulus with flow obstacles. Despite of the significant effort exercised in the last few decades, the physical phenomena that govern the post-dryout heat transfer in the vicinity of obstacles is still not well understood. In particular, recent investigations (Anglart and Persson, 2007) indicate that the current prediction methods may significantly under-predict the effect of obstacles (such as spacer grids in nuclear fuel assemblies) on post-dryout heat transfer. The objectives of present experiments are to obtain a new insight into the obstacle influence on post-dryout heat transfer and to build a detailed database in order to improve current models used in the post-dryout calculations. The reference experiments are performed in the annulus with pin spacers. The influence of the pin spacers on the post dryout heat transfer was expected to be insignificant. Two possible situations are investigated: occurrence of the CHF on the inner rod and occurrence of the CHF on the outer tube. After reference cases studies, the test section is dismantled and two cylindrical spacers are mounted on the inner rod. The experimental results obtained in these cases are compared with the reference cases.
The experiments are carried out in the High-pressure Water Test (HWAT) loop in the Nuclear Reactor Technology Division of the Royal Institute of Technology in Stockholm, Sweden. The HWAT experimental facility is designed to operate at pressures up to 25 MPa, power up to 1 MW and water mass flow rate up to 1.2 kg/s. The test section consists of two-side heated annulus (12.7x24.3) mm, with total heated length equal to 3650 mm. For the reference cases pin spacers are employed to keep the rod in the central location in the annulus. They are positioned at five levels evenly displaced from the annulus inlet. In the experiments with obstacles, two ring-shaped obstacles are inserted into the annulus to simulate the spacer grids of BWR fuel assembly.
The experimental investigations are performed with both (with and without obstacles) types of test sections in a wide range of the operational conditions: mass flux (500-2000) kg/(m2s), inlet subcooling (10-40) K and system pressure (5-7) MPa, with focus on typical operating conditions in Boiling Water Reactors. The temperature data has been collected using 88 thermocouples placed in the test section: 40 thermocouples are placed axially inside of the rod to measure the rod wall temperature, 40 thermocouples are placed outside of the tube to measure the tube wall temperature, and finally, 8 thermocouples are placed azimuthally before and after the last spacer to determine the azimuthally temperature distribution on the tube wall at that position. The data has been collected using a data logger device and the Labview program. The full paper will contain a detailed description of the experimental procedure and a short analysis of the experimental results.
34 Nuclear Energy f0r New Eur0pe 2010
References
[1] Anglart, H. and Persson, P., “Experimental Investigation of Post-Dryout Heat Transfer With Spacers”, Volume 33, Issue 8, August 2007, Pages 809-821,Int. J. of Multiphase Flows.
_406 ______
A Model for Droplet Entrainment in Horizontal Stratified Flow
F. Henry, Y. Bartosiewicz Université catholique de Louvain UCL, Louvain School of Engineering EPL Place du Levant 2, B – 1348 Louvain-La-Neuve, Belgium [email protected], [email protected]
This work investigates the ability of the CATHARE 3 system code to predict droplet entrainment in a mechanistic approach based on development of interfacial instabilities in order to replace semi- empirical models.
Steam water stratified flows occurred in the hot leg of nuclear reactor in some local transients. The droplets entrainment results of the growth of interfacial waves at the interface gas-liquid. In horizontal pipes, the ability of system codes to predict this entrainment phenomena and the deposition of these entrained droplets is required to effectively calculate the mass transfers between phases. The modeling of droplets which will be entrained to steam generators is extremely important to determine the pressure variations in the core and then for subsequent core reflooding process. This work is an attempt to develop a model based on a underlying physics of wave appearance, breaking and droplet entrainment.
A first step was already performed as a flow linear stability analysis in order to calculate wavelengths of interfacial instabilities from flow characteristic. The dispersion equation of interfacial waves was derived and compared from results from literature. In this paper, the present study completes this stability analysis and proposes a model for droplet entrainment which could be implemented in CATHARE 3 system code in a forthcoming step . This model is developed based on a force balance applied at wave crests in order to determine waves breaking. Effects of flow regimes and pipe dimensions are investigated and compared to results from literature.
_407 ______
Loss of External Load Analysis using RELAP5/MOD3.3 Patch 03 Computer Code
Andrej Prošek “Jožef Stefan” Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected]
After leakage in 2008 Krško nuclear power plant considers to remove resistance temperature detectors (RTD) bypass. Fast acting RTD/thermowell response time is slower than response of direct immersion RTD, but there is no loop transport or thermal lag. Depending on the delay of fast acting RTD, new analysis may be needed or not. Faster fast acting RTD do not require new analysis, but they are more costly. The aim of this study was to investigate the influence of measurement delay on OTΔT trip setpoint calculation during loss of external load analysis, which was the transient of interest. To get 35 Nuclear Energy f0r New Eur0pe 2010 reactor trip on OTΔT in the original updated safety analysis report (USAR) analysis the reactor trip on high pressurizer pressure was disabled by assumption. Besides varying the delay time also the influence of different conservative assumptions used in the USAR analysis were studied. For simulations the latest RELAP5/MOD3.3 Patch 03 thermal hydraulic computer code was used. The verified standard RELAP5/MOD3.3 input model from 2008 (cycle 23) was delivered by Krško nuclear power plant. Due to limitation of RELAP5 to calculate the departure from nucleate boiling ratio (DNBR), the calculation of OTΔT (OPΔT) setpoint was investigated as a function of temperature measurement delay. First comparison with the original USAR analysis was done to show that main plant parameters agree with USAR calculation and that the quantitative differences are understood. Then sensitivity analysis was performed. The results showed that when the temperature measurement delay is rather small (e.g. ± 2 seconds), also the trip on OTΔT is similarly delayed. After reactor trip the main parameters like pressurizer pressure and level, hot and cold leg temperature and secondary pressure start to decrease. Because the trip times are delayed, also the time trends are time shifted. Before reactor trip all these parameters were increasing except pressurizer pressure, which increase was limited by pressurizer relief valves opening. This means, the larger the measurement delay, the higher is the average temperature of the reactor coolant system because of delayed reactor trip on OTΔT. The results also showed that for small delays the relations are linear. Having the DNBR trend from original USAR calculation, one can thus at least qualitatively assess how measurement delay time influence the DNBR during loss of external load.
_408 ______
Direct Numerical Simulation of Heat Transfer at Very Low Prandtl Number:Application to Convection in Lead-Bismuthflows
1 1 1 Laurent Bricteux , Matthieu Duponcheel , Yann Bartosiewicz 1 Universit´e catholique de LouvainInstitute of Mechanics, Materials and Civil Engineering (iMMC)Place du Levant 2, 1348 Louvain la Neuve, [email protected]
Liquid Metal Reactors (LRM) represent a promising technology for achieving the various criteria required to be certified as a GENIV concept. For those reactors, two coolants are envisaged: the sodium and a lead-bismuth eutectic (LBE). The Prandtl number of such fluids is very low, for instance
Pr0.01for LBE and Pr0.001for sodium at operational conditions. The most common modeling approach for the simulation of turbulent heat transfer is based on the the Reynolds analogy by using the so-called turbulent Prandtl number (Prt). As a result the eddy-diffusivity concept is applied to heat transport, however, proven to be not valid for liquid metal with low molecular Prandtl number. Therefore another approach will be considered in the THINS EU (FP7) project. It consists in solving 2 transport equations for θ and its dissipation rate θ, and using a non linear heat flux model. In this regard, Direct Numerical Simulation (DNS) should be performed to provide a sufficient database at relevant Reynolds and Prandtl numbers. This paper presents recent developments of DNS for liquid metals with a focus on local heat transfer behavior. In order to capture such complex physics, high quality numerical methods are required, with negligible numerical dissipation (i.e., methods that con- serve energy in absence of viscosity) and with low dispersion errors (to properly transport the turbulent structures). A fourth order energy conserving finite difference parallel code is used to satisfy
36 Nuclear Energy f0r New Eur0pe 2010 this criteria. Moreover, the numerical requirements to perform direct numerical simulations (DNS) of turbulent flows for liquid metals are very stringent. Indeed, the conjunction of a high Reynolds number and a very low Prandtl number leads to severe constraints on the time step and the grid size. This paper will first expose in a clear and synthetic way what are the numerical constraints to proper simulate such a flow in the case liquid metal channel. In the second part, numerical predictions of two flows of interest will be presented. The first case is the heat transfer in a turbulent channel flow at Reτ=180(based on the friction velocity) and Pr=0.01. The turbulent channel flow with heat transfer is a test case which is well documented in the literature from moderate to low Prandtl number
Pr10→Pr0.1(see e.g. the work of Kasagi et al. [1] and Kawamura [2]). The second case deals with the onset of thermal instabilities in the Rayleigh-Benard problem at high Rayleigh number and in the case of LBE with Pr=0.01.
References
[1] Kasagi, N. and Iida, O. 1999, Progress in direct numerical simulation of turbulent heat transfer, Proceedings of the 5th ASME/JSME Joint Thermal Engineering Conference, In CD-ROM, March 15- 19, San Diego, California.
[2] H. Kawamura, K. Ohsaka, H. Abe, K. Yamamoto, DNS of turbulent heat transfer in channel flow with low to medium-high Prandtl number fluid, International Journal of Heat and Fluid Flow, Volume 19, Issue 5, October 1998, Pages 482-49
_409 ______
Controlling Valve Fluid Jet Energy to Eliminate Vibration
Tobias Zieger Principal Engineer IMI Nuclear Im Link 11, 8404 Winterthur, Switzerland [email protected]
Jay Faramarzi VP Technology & Product Development CCI Mark A. Hollerbach, Mgr Engineering CCI Herbert L. Miller, Consultant
Retrofitting the valve trim of problem valves with a trim that provides velocity control of the jets exiting the trim has been implemented on almost 500 valves in the field. The retrofit data has been sorted to examine only those problem valves where the user noted a vibration problem. This population represented 116 valves. A valve associated with a vibration problem may also have other problems such as controllability, noise and erosion to name a few. A review of this retrofit experience implies a strong link between controlling the energy level of the fluid jets exiting the valve trim and problems of vibration. Every time an original trim was replaced with a velocity control trim the vibration problem was eliminated. The dominate change was a much lower fluid jet energy exiting the trim. The energy reduction ranged from a high of 155 times and lower depending upon the magnitude of the original trim fluid jet energy and averaged a reduction of 15 times less energy.
37 Nuclear Energy f0r New Eur0pe 2010
The valve fluid jet energy levels for all trims are many times higher than the pipe flow energy levels. This is a result of the smaller flow area associated with the valve flow passage. The valve design is much more ruggedly designed that the piping and thus the higher energy levels do not create a problem within the valve. Dynamic pressure of piping fluids are usually less than 70 kPa (10 psi) whether the flowing medium is liquid or gas and regardless of any valve design in the system. In some of the valve designs with vibration problems the jet dynamic pressure, kinetic energy density, was as high as 9.5 MPa (1400 psi).
The vibration retrofit data includes pipe sizes that range from 4 to 24 inch, outlet pressures from a vacuum to 20 MPa (2900 psi) and fluids of steam, water and hydrocarbons. There were 64 valves handling liquids and 52 handling gases. The population of valves represented 35 different valve designs from 9 different manufacturers.
_410 ______
Comparative Reliability Analysis for Two Different Designs of Residual Heat Removal System (RHRS) and Containment Spray System (CSS) of IR-360 Nuclear Power Plant
Faramarz Yousefpour1, Kaveh Karimi2,1, Hamid Soltani1 1Management of Nuclear Power Plant Construction Company (MASNA Co.), Tehran, Iran 2Islamic Azad University, Science and Research Branch, Tehran, Iran Email address: [email protected]
IR-360 nuclear power plant, a 2-loop 360 MWe PWR, is under design in MASNA Company. Design of Residual Heat Removal System (RHRS) and Containment Spray System (CSS) is already accomplished. RHRS is a stand alone system with two trains; each of which consists of one RHRS pump and one heat exchanger shared with Low Pressure Safety Injection System (LPSIS) for long term cooling in LOCAs (Loss of Coolant Accidents). CSS is also a stand alone system with two trains consisting of their own pump, heat exchanger and spray nozzles.
Besides, EPRI Utility Requirements Document (URD) requires RHRS interconnected with CSS in a way that both systems could use each other pumps. This configuration, i.e. CS/RHR system, permits maintenance to be performed on either an RHRS or CSS pump during normal operation or refueling without any time restrictions and permits the RHRS pumps to back up the CSS pumps during long- term post-LOCA operations.
In this study both different designs are analyzed in order to assess the advantages of EPRI URD required design rather than the already design of IR-360 RHR & CS systems.
First failure mode and effect analyses are performed for both designs. Then comparative reliability analyses are performed by fault tree analysis approach using SAPHIRE code.
Parameters are treated as distributions in the analyses, therefore final reliability of each design are provided with their uncertainty bounds.
Results showed that the order of magnitude of reliabilty of second configuration is nearly doubled in comparison with first configuration. This means that the second configuration is safer than the first one. Moreover using the second configuration would give the chance for maintenance of trains during the plant operation.
38 Nuclear Energy f0r New Eur0pe 2010
Importance measure analyses are also performed for both designs, in which Fussll-Vessely importance, risk reduction ratio, risk increase ratio and Birnbaum importance factors are mainly considered.
Common cause failure analysis is also performed in this study.
_411 ______
Preliminary CFD Study of Flow Oscillations in Parallel Channels Using the Volume of Fluid (VOF) Method
Davide Papini, Giovanni Pastore, Antonio Cammi Politecnico di Milano – Department of Energy CeSNEF – Nuclear Engineering Division Via La Masa, 34 – 20156 Milano, Italy [email protected]
Computational Fluid-Dynamics (CFD) codes are increasingly adopted for the analysis of critical issues concerning the operation and safety of various industrial applications, since they provide a detailed insight into the physical mechanisms involved. For instance, prediction of flow oscillations in two- phase systems is of interest to the design of steam generators and boiling water nuclear reactor cores.
In this work, a preliminary CFD study is presented, aimed at investigating the transient behaviour of a two-phase system experiencing flow oscillations. The addressed system is composed by two parallel channels, and a boundary condition involving the same pressure drop across the channels is provided by the connection with common upper and lower headers, forming a closed 2-D domain filled with saturated water and saturated steam. Calculations are performed using the FLUENT code, and the Volume Of Fluid (VOF) method, which is a surface-tracking technique designed for two (or more) immiscible fluids, is selected as multiphase flow model. Liquid phase and vapour phase are perfectly separated, heat generation and heat transfer in the system are neglected, and the onset of oscillations is provided by imposing unbalanced initial conditions in the gravitational pressure drop term (different positions of the water-steam interface location in the two channels). The results demonstrate the capability of the VOF method in tracking the interface motion.
The system behaves as a damped non-linear oscillator, and the performed simulations allow evaluating the oscillation frequency and decay ratio, as well as the sensitivity to working conditions (e.g., operating pressure), which are known to play a relevant role in this kind of mechanism. Moreover, a picture of the fluid-dynamics involved is obtained, showing a close relationship between the oscillatory behaviour of the system and the time evolution of fluid vortices, forming at the channel inlet sections. Fluid-dynamics details give possible explanation to some peculiar findings in terms of channel pressure distributions, which are in agreement with other numerical studies available from literature.
39 Nuclear Energy f0r New Eur0pe 2010
_412 ______
Analysis of the Two-phase Flow Phenomena at PGV Test Facility
Yury Parfenov, Vladimir Melikhov, Oleg Melikhov, Alexey Nerovnov Electrogorsk Research and Engineering Center on Nuclear Power Plants Safety Saint Constantine st., 6, Electrogorsk, Moscow region, 142530, Russia [email protected], [email protected], [email protected]
New design of nuclear power plant (NPP) with pressurized water reactor «NPP-2006» is developed in Russia. It represents evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in «NPP-2006». The equalization of steam release in this steam generator can be provided by use of the submerged perforated sheet (SPS) with nonuniform perforation. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the effect of the nonuniform degree of the SPS perforation on the distribution of the thermal-hydraulic parameters in steam generator and to obtain data for the thermal-hydraulic code verification. The PGV test facility consists of a slice model of the upper part of the PGV-1000MKP steam generator installed in the high pressure vessel. The description of the PGV test facility and the most important experimental results obtained at the facility are presented in the paper. The experimental results were used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest and posttest calculation results are presented in the paper.
_413 ______
Numerical Investigation on Boiling Channel Instabilities by Imposing Constant Pressure Drop Boundary Condition Via a Large Bypass
Marco Colombo, Antonio Cammi, Davide Papini, Marco Ricotti Politecnico di Milano – Department of Energy CeSNEF – Nuclear Engineering Division Via La Masa, 34 – 20156 Milano, Italy [email protected]
Density Wave Oscillations (DWOs) are probably the most common type of instabilities affecting vapour generation in boiling systems. They are “dynamic type” instabilities, which result from multiple feedback effects between the flow rate, the vapour generation rate and the pressure drops in the boiling channel. Several experimental studies on density wave instabilities have been reported in the open literature since the 60s-70s, due to the great importance that instability concern has for the operation and safety of various industrial applications, including steam generators and boiling water nuclear reactor cores.
A constant pressure drop is the proper boundary condition which can excite the fundamental mechanisms leading to the appearance of DWOs in a single channel. Therefore, several experimental works were conducted using a large bypass pipe, to impose the mentioned boundary condition on the boiling channel. As a matter of fact, the mass flow rate is generally forced in an experimental apparatus by a feedwater pump, instead of being driven by a pressure difference imposed across the
40 Nuclear Energy f0r New Eur0pe 2010 channel. A large bypass tube is hence necessary to maintain the constant pressure drop condition on the single heated channel.
In this frame, the described experimental layout is investigated by means of the RELAP5/MOD3.3 code: a single heated pipe is connected to a large bypass by means of two branches, which also permit to account for local pressure losses. The simulation data are clustered in dimensionless stability maps, generally adopted in such stability investigations.
In the recent years, many numerical studies devoted to density wave phenomena in a single heated channel have been presented. Among them, Ambrosini and Ferreri [Analysis of Basic Phenomena in Boiling Channel Instabilities with Different Flow Models and Numerical Schemes, Proceedings of ICONE 14, July 17-20, 2006, Miami, Florida, USA] demonstrated the capability of the RELAP5 code to detect the onset of instability, when a single pipe is connected to two inlet and outlet plena kept at imposed pressures.
The stability maps obtained in this work with the bypass layout show a good agreement when compared to the results of the single pipe layout with imposed pressures. In particular, the results are satisfactory for a bypass ratio large enough to accommodate the instabilities of the heated channel without disturbing the constant pressure boundary. The influence on stability of the bypass ratio is also investigated.
The results represent a contribution to the assessment of the code capability to detect the onset of two- phase flow instabilities in a boiling channel.
_414 ______
Screening Methodology for the Evaluation of PTS scenarios in the Symptom based Emergency Operating Procedures
Ilijana Iveković, Tomislav Bajs ENCONET d.o.o. Miramarska 20, 10 000 Zagreb, Croatia [email protected], [email protected]
Ivica Bašić APOSS d.o.o. Repovec 23B, 49210 Zabok, Croatia [email protected]
Integrity of the reactor pressure vessel must be preserved throughout the plant operation as well in the postulated accidental conditions. Consistently, the acttions that operators are instructed to perform in order to mitigate accidents should not adversely affect integrity of the reactor pressure vessel. Special consideration in this frame is dedicated to the Pressurized Thermal Shock transients (PTS), overcooling transients (interaction of cold safety injection water and hot vessel wall) with high pressure in reactor coolant system, which can cause damage to the reactor pressure vessel.
PTS screening methodology was used according to determined criteria to select the most severe PTS scenarios caused by operator actions for two loop PWR nuclear power plant. The aim was to provide bounding analysis for the additional verification of the symptom based Emergency Operating Procedures (EOP). In order to bound asymmetrical behaviour of the two loop plant, transients with the most adverse screening criteria were further analyzed with vertically split reactor vessel model and screening criteria were applied to the results obtained using RELAP5 computer code and split reactor
41 Nuclear Energy f0r New Eur0pe 2010 pressure vessel model Additionally, on the basis of requirements from the 10 CFR 50.61, methodology accepted by USA NRC, an assessment was performed to project and assesses the values of PTS temperature (RTPTS) for the analysed PWR nuclear power plant.
_415 ______
Comparison of R5G Coupled Code and Classical “Two-steps” Containment Calculation
Davor Grgić, Vesna Benčik, Siniša Šadek Faculty of Electrical Engineering and Computing, University of Zagreb Unska 3, 10000 Zagreb, Croatia [email protected], [email protected], [email protected]
In the case of interaction of the reactor system and containment, usually two independent runs are performed. First the mass and energy release rates (MER) are calculated by a system code using conservative assumption of low containment back pressure, and then containment response is calculated using the calculated MERs as a boundary condition. This approach is usually adequate for present day Light Water Reactors (LWRs), due to low importance of the interaction between containment and reactor coolant system during Design Bases Accidents (DBAs), typically Loss of Coolant Accident (LOCA) and Steam Line Break (SLB). However, in new advanced passive systems such as in International Reactor Innovative and Secure (IRIS) and Economic Simplified Boiling Water Reactor (ESBWR), the LOCA phenomenology can be significantly different, and interaction of the coolant system and containment becomes more important, so as to require an evaluation model where containment and reactor coolant system are analyzed in a coupled way.
Even though usage of coupled system and containment code is not necessary for present days LWRs, from point of view of prediction accuracy, once when coupled code is developed, it can make safety analyses easier and more complete.
The coupled code R5G is a result of explicit coupling of system code RELAP5/mod3.3, and containment code GOTHIC, done at the University of Zagreb.
In the paper the calculations done in frame of NPP Krsko Full Scope Simulator (KFSS) verification were performed using both, classical “two-steps” approach, and single run using coupled code. The most important system and containment variables were compared and reasons for differences were explained for SLB and stuck open PORV calculation cases.
_416 ______
The Integral Analysis of 40 Mm Diameter Pipe Rupture in Cooling System of Fusion Facility W7-X With Astec Code
Tomas Kačegavičius Lithuanian Energy Institute Breslaujos g. 3, LT-44403 Kaunas, Lithuania [email protected]
Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental facility of
42 Nuclear Energy f0r New Eur0pe 2010 stellarator type, which is currently being built in the Greifswald branch institute of IPP, Germany. W7- X shall demonstrate that in future the energy could be produced in such type of fusion reactors. The safety analysis is required before the operation of the facility could be started. Rupture of 40 mm diameter pipe, which is connected to the target to ensure heat removal from the vacuum vessel in case of no-plasma operation mode “baking” is one of the design basis accidents. “Baking” is the mode of facility operation during which the vacuum vessel structures are heated to the temperature required for plasma ignition. This accident was selected for the detailed analysis using integral code ASTEC, which is developed by IRSN (France) and GRS mbH (Germany).
This paper presents the integral analysis of W7-X response to a selected accident scenario. At first there is described development of the model of the main cooling circuit and “baking” circuit for CESAR module, and model of vacuum vessel for CPA module of ASTEC code. Later the results of performed analysis are presented. There were analysed two cases: 1) rupture of pipe connected to the upper target and 2) rupture of pipe connected to the lower target. The results of analysis showed that more water is discharged in case of rupture of pipe connected to the lower target. In both cases the primary and secondary steam production leads to a rapid increase of pressure in the plasma vessel and in 22 s the pressure increases to the set point of safety valve opening, which prevents further rise of pressure.
_417 ______
Numerical Simulation of Turbulent Subcooled Boiling Flow in a Rectangular Channel
Boštjan Končar “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
Carlos E. Estrada-Perez, Yassin A. Hassan Texas A&M University College Station, Texas, 77843, USA [email protected], [email protected]
This work is related to the collaborative project of the 7th European framework program NURISP (NUclear Reactor Integrated Simulation Platform). One of the goals of the NURISP project is also to assess and improve the simulation capability of the three-dimensional two-fluid codes for prediction of local boiling flow processes.
The flow boiling is strongly influenced by the near-wall mechanisms (bubble nucleation, bubble growth, sliding, detachment, bubbles merging on the wall, rewetting after detachment, etc.). The structure of boiling flow is typically non-homogeneous and is further influenced by the core-flow processes, such as bubble interactions, flow turbulence, condensation in the subcooled core flow. For computational predictions of realistic bubbly flows, the use of two-fluid Eulerian approach is the best available choice. To validate the two-fluid models, the use of reliable channel flow experiments, which provide quality measurements of local distributions of flow parameters, is indispensable.
In this paper the new set of boiling flow experiments, performed recently at the Texas A&M University, is used to test the prediction capabilities of the two-fluid modelling approach. Subcooled boiling flow experiments were carried out in a vertical square channel with a single heated wall. The refrigerant Novec-7000 was used as a working fluid. Experiments were performed at ambient outlet
43 Nuclear Energy f0r New Eur0pe 2010 pressure at three different mass flow rates and 13 different heat fluxes ranging from 0 up to 64 kW/m2. Particle Image Velocimetry (PIV) technique was used to measure time-averaged velocities, turbulence intensities and Reynolds stresses. These experimental data are particularly useful for validation of two- fluid turbulence models used for modelling of the two-phase boundary region near the heated wall.
44 Nuclear Energy f0r New Eur0pe 2010
Multiphyiscs
45 Nuclear Energy f0r New Eur0pe 2010
_501 ______
Implementation of New Simulation Capabilities in RELAP5/PARCS V2.7 Coupled Codes
R. Miró1, C. Pereira2, T. Barrachina1, G. Verdú1, J. C. Martínez-Murillo3
1Chemical and Nuclear Engineering Department Polytechnic University of Valencia Camí de Vera s/n, 46021 Valencia, Spain
2Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais Av. Antonio Carlos 6627, 31270-901 Belo Horizonte, Brazil
To efficiently characterize realistic transients, as the Reactivity Insertion Accidents (RIA), using coupled neutronic-thermal-hydraulic 3D best estimate system codes, like RELAP5/PARCS v2.7 coupled code, it is necessary to introduce some improvements in simulations by adding the capability of control rod movement and boron injection by means of RELAP5 control variables, with the aim of being able to analyze dynamically asymmetric transient accidents in a nuclear power reactor, like RIA, reproducing all control systems present in commercial reactors.
In actual neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; control rod movements are pre-programmed with simple instructions introduced in the input deck before the beginning of the calculation.
In order to simulate a boron injection/dilution transient it necessary to modify the neutronic code to be able to read and interpolate between the borated and non-borated cross-sections sets. Besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution.
This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. With these modifications, control rods can be moved automatically, activated by the RELAP code control system, and also they can depend on signals related to the reactor activity, like pressure, fuel temperature or moderator temperature, etc., improving the realism of the calculation and widening the simulation possibilities. RELAP5 calculates the boron concentration in each node of the channels representing the reactor core, sending this information to the PARCS neutronic code.
As practical qualification, a RIA transient and a boron injection transient for TRILLO NPP are presented.
46 Nuclear Energy f0r New Eur0pe 2010
_502 ______
The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit
Peter Hermansky, Marian Krajčovič VUJE, Inc. Okružná 5, 918 64 Trnava, Slovakia [email protected], [email protected]
The functions of the WWER 440/V213 reactor internals are to support the core, to hold the fuel assemblies in place, to direct coolant flow, to hold and protect emergency control assemblies in normal operation conditions and accidents conditions.
In the case of a LOCA accident it is assumed rapid “guillotine” break of one of the main coolant pipes and rapid depressurization of the primary circuit. The pressure wave propagates at the speed of sound, enters the reactor pressure vessel and causes unsymmetrical loads of the reactor vessel internals. These loads are only important in the initial phase of the accident – within a time interval of tenths of second after the occurrence of the accident. After this interval, the pressure comes to balance and the dynamic load of the reactor vessel internals disappears.
This paper presents results of the numerical simulation of the WWER440/V213 reactor vessel internals (RVI) dynamic response to LOCA accident assumed in the hot and cold leg of the primary circuit.
The global thermo-hydraulic calculation of the LOCA accident was performed by means of the RELAP5/Mod 3.2.2 code by using the six loop model of the WWER440 reactor cooling system. The RVI finite element model was created by MSC.Patran and dynamic response was solved using MSC.Dytran code. The model consists of reactor vessel internals (Lagrangian solid elements) and water coolant (Euler elements) inside the reactor. Arbitrary Lagrangian Eulerian coupling was used for simulation of the fluid-structure interaction. The calculation assumes no phase change in the water.
The nuclear power plant safety analysis guidelines define basic requirements and conditions for accident analyses. The most important acceptance criteria for the reactor internals demands that the movement of the emergency control assemblies under all operating conditions including accident is ensured.
The numerical simulation of the WWER440/V213 reactor internals response to a LOCA accident in the hot and cold leg showed that the acceptance criteria for RVI is fulfilled and required NPP safety standards are satisfied.
47 Nuclear Energy f0r New Eur0pe 2010
_503 ______
Fracture Mechanics Analysis in a Pressurized Heavy Water Reactor Vessel during LOCA scenario
Dino A. Araneo, Giuseppe Agresta, Francesco D’Auria GRNSPG, DIMNP, University of Pisa Via Livornese 1291, 56122 Pisa, Italy [email protected], [email protected], [email protected]
The Reactor Pressure Vessel (RPV) has long been considered one of the most reliable components in Pressurized Water Reactor (PWR). Nowadays, a general target for the countries that produce nuclear energy is to extend the operation life of existing plants. From this point of view, the RPV is one of the major components that may limit the useful life of the nuclear plant. The risk for the RPV structural integrity is connected to the presence of a flaw of sufficient size, a high level of embrittlement due to radiation damage, and the occurrence of a thermal-hydraulic transient inducing strong stresses in the vessel wall.
Severe loading conditions are produced during a Pressurized Thermal Shock (PTS) event, in which an overcooling may induce strong thermal stresses while the internal pressure can be maintained at high level or the system can be repressurized during the transient. Such conditions are generated in a LOCA transient during the emergency injection.
In recent years, important progresses have been made in the development of analysis methods and tools for the best estimation of the thermal and pressure loads on the vessel wall. In this direction, the US-NRC published in 2007 a document aimed at reviewing the rules adopted in PTS analysis, established in the 1980s, containing significant conservatisms, for a Best Estimate (BE) approach combined with uncertainty assessment.
In this paper, the methodology for Fracture Mechanics analysis developed at University of Pisa aimed to perform parametric analysis assuming various shapes and locations of the flaw is applied to a Pressurized Heavy Water Reactors (PHWR) during a LOCA scenario. Four steps can be identified starting from the thermal hydraulic analysis of the Nuclear Power Plant (NPP) behaviour with Relap5- 3D© in order to calculate the cooling loads of the Emergency Core Coolant Systems (ECCS). The second step is the analysis by mean a CFD code (CFX) of the mixing phenomena occurring in the Down-Comer (DC) and the calculation of the thermal load on the RPV internal surface. The third step is represented by the evaluation of the stresses inside the RPV wall by mean a FE code (Ansys) under the thermal and pressure loads calculate in the previous steps. The last step is represented by the calculation of the Stress Intensity Factor (SIF) KI by mean the Weight Function method and the comparison with the critical SIF KIc of the material, once the stresses inside the undamaged RPV wall are known.
The goal of this work is the evaluation of the safety margin for the operation of the RPV, adopting a BE approach in all the steps of the analysis. This result will be compared with the one obtained with the application of the ASME XI criteria for the KI evaluation with the aim to show that the BE approach leads to a big safety margin.
48 Nuclear Energy f0r New Eur0pe 2010
_504 ______
TRANSURANUS Verification against CNEA’s PHWR MOX Experiments, from IFPE Database
Rozzia D., M. Adorni, A. Del Nevo, F. D’Auria University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG), Via Diotisalvi 2, 56122 Pisa [email protected], [email protected], [email protected], [email protected]
Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. The comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident conditions are part of the defense in depth concept.
In this connection, OECD NEA sets up the public domain database on nuclear fuel performance experiments – International Fuel Performance Experiments (IFPE) database, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for models development and codes validation. The CNEA’s PHWR MOX experiments belong to this database.
These experiments were carried out in the High Flux Reactor (HFR) of Petten, Holland. They involve six MOX rods prepared and controlled in the CNEA’s “alpha” Facility.
The first rod was used for destructive pre-irradiation examinations. The second one was a “pathfinder” rod irradiated for about 100h plus a final ramp with the aim to adjust systems in the HFR.
Two additional rods were doped with iodine and irradiated for about 15 days plus final ramps. The iodine concentration simulates a burnup of 15 MWd/kgU.
Finally the BU15 experiment was performed with the last two rods. The objective of this experiment was to verify the fabrication processes and the study of the fuel behavior with respect to cladding failure due to Stress Corrosion Cracking (SCC) under Pellet Cladding Interaction (PCI) conditions. These rods were irradiated until 15 MWd/kgU and then one them experienced a final ramp.
The code TRANSURANUS version “v1m1j09” is assessed against the database CNEA’s PHWR MOX in order to verify the capability of the code in predicting the failures due to stress corrosion cracking and the associated phenomena and processes prior and after ramps. The results presented include the complete set of simulations of all rods irradiated in the HFR reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of code options and boundary conditions on results. A comparison with the results obtained with BACO code by “Comisión Nacional de Energía Atómica”, Bariloche is also presented.
49 Nuclear Energy f0r New Eur0pe 2010
_505 ______
Reactivity Initiated Accident Analysis of the HPLWR Three Pass Core with Ascending Gap Flow using the KIKO3D-ATHLET Code
György Hegyi, András Keresztúri, Csaba Maráczy, István Trosztel MTA KFKI Atomic Energy Research Institute Budapest, Hungary H-1525 [email protected], [email protected], [email protected], [email protected]
The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of safety analysis of reactivity initiated accidents (RIA) within the framework of the European funded “HPLWR Phase 2” and the Hungarian sponsored “NUKENERG” projects.
As the target average outlet temperature (500 oC) and the maximum foreseen cladding temperature (630 oC) are very close to each other, the 3-pass core concept was proposed. The hot spots in this proposal can be potentially eliminated by multiple flow of coolant through the active core with mixing after each passing. Because of the possible unwanted bouyancy effects in case of descending assembly gap flow of moderator, the flow path of moderator water was modified from the downward gap flow to upward gap flow which results in the modification of reflector regions to provide space for the downstream water. The increased mass flow rate of moderator channels provides better neutronic moderation for the thermal core influencing the power distribution. The reactivity coefficients of HPLWR with respect to fuel temperature and water density show similarity to current pressurized water reactors at normal operating conditions but the power response of the core to reactivity changes is quite different. The water feedback plays more important role than in PWR reactors. The modified flow path causes an extra time delay for the important moderator reactivity feedback in case of RIA type events. The dynamic core behaviour of the reactor corresponding to the above concept was investigated in cases of the following RIA events:
• Uncontrolled absorber group withdrawal
• Control rod ejection
• Loss of feedwater heating
• Control rod malfunction
The acceptance criteria are fulfilled, however, in spite of the relatively strong feedback, in some cases the hot channel temperatures are not far from the limits because of the significant perturbation of the local power, which points out the necessity of the RIA analyses.
50 Nuclear Energy f0r New Eur0pe 2010
_506 ______
Uncertainty and Sensitivity Analysis in the Neutronic Parameters Generation for BWR and PWR Coupled Thermalhydraulic-Neutronic Simulations
J. Juanas1, T. Barrachina2, R. Miró2, R. Macián1, G. Verdú2 1Lehrstuhl für Nukleartechnik Technical University Munich Boltzmannstrasse 15, 85747 Garching, Germany 2Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM) Universitat Politècnica de València (UPV) Camí de Vera s/n, 46021 Valencia, Spain
This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional BWR core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA).
The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal- hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE.
The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system´s behaviour, uncertainties from both aspects should be included and jointly propagated.
The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether-Wilks formulas.
A number of simulations equal to the sample size is carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.
51 Nuclear Energy f0r New Eur0pe 2010
Materials and Structural Integrity
52 Nuclear Energy f0r New Eur0pe 2010
_601 ______
Corrosion Experience with the Secondary Side of CANDU Steam Generator Tubesheet
Dumitra Lucan Institute for Nuclear Research, Pitesti, Romania,
Nearly all of the steam generator fouling processes and the related degradations were attributed to secondary side water chemistry conditions and excursions, many of which resulted from condenser cooling water ingress.
The investigation of the structural materials corrosion in correlation with the water chemistry, as well as the impurities and corrosion products concentration and deposition and their removing from the CANDU steam generators is a very active field and both the experimental works and the understanding of the mechanisms involved are submitted to some rapid changes and permanently open to the research. To provide information about the corrosion behaviour of the structural materials from CANDU steam generators under normal and abnormal conditions of operation and to identify the failure types produced by corrosion we performed a lot of corrosion experiments. These experiments consisted in chemical accelerated tests, static autoclaving and electrochemical investigations.
The purpose of this experimental research consists in the assessment of corrosion behaviour of the tubesheet material, carbon steel SA 508 cl.2, at normal secondary circuit parameters (temperature- 2600C, pressure-5.1MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment – AVT).
The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.
The gravimetric method, optical metallographic microscopy as well as electrochemical measurements have been used to evaluate the corrosion behavior of the tubesheet material.
_602 ______
Towards Modeling Intergranular Stress Corrosion Cracks on Grain Size Scales
Igor Simonovski, Leon Cizelj Jožef Stefan Institute, Reactor Engineering Division Jamova cesta 39, 1000 Ljubljana, Slovenia [email protected], [email protected]
Development of advance grain size scale models has so far been mostly limited to simulated geometry structures such as for example 3D Voronoi tessellations. The difficulty came from lack of non- destructive techniques for measuring the microstructures. With recent development of novel experimental technique of X-ray diffraction contrast tomography (DCT) the researchers were given a unique tool for non-destructively measuring specimen’s microstructure. The DCT method is being developed by the Materials Performance Centre of The University of Manchester.
53 Nuclear Energy f0r New Eur0pe 2010
The paper present a model of intergranular stress corrosion cracks, based on a measured microstructure of a 400 μm stainless steel wire. The wire is composed of 362 grains and over 1600 grain boundaries. A finite element model of the wire is presented. Grain shapes are reconstructed from the experimental data and assigned measured crystallographic orientations. Grain boundaries are explicitly modelled by employing cohesive elements with zero physical thickness. Grain boundaries are classified according to the crystallographic orientations of the neighbouring grains into corrosion- resistant (strong) and corrosion-susceptible (weak) grain boundaries. The development and growth of stress-corrosion cracks is enabled by using cohesive elements with incorporated damage properties. The preliminary results of development and growth of intergranular stress corrosion cracks pertaining to the isotropic material model of the grains is presented in the paper.
_603 ______
Observation of Long-term Corrosion at Nuclear Power Plant Bohunice (Slovakia)
Vladimir Slugen1, Jozef Lipka1,3, Julius Dekan1, Jarmila Degmova1, Ignac Tóth1, Pavol Šeliga1 and Ivan Smieško2 1Department of Nuclear Physics and Technology, Slovak University of Technology Bratislava, Ilkovicova 3, 812 19 Bratislava, Slovakia; [email protected] 2NPP Jaslovské Bohunice, SE, a.s., Slovakia. 3Slovak Institute of Metrology, Bratislava, Slovakia
Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original “Bohunice” design in period 1994-1998. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Mössbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Mössbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filters deposits. The corrosion of new feed water pipelines system (from austenitic steel) in combination to innovated operation regimes goes dominantly to magnetite. The hematite presence is mostly on the internal surface of steam generator body and its concentration increases towards to top of the body. In the results interpretation it is necessary to consider also erosion as well as scope and type of maintenance activities. The long-term study of phase composition of corrosion products at VVER reactors is one of precondition for the safe operation over the projected NPP lifetime.
_604 ______
Modular 3-D Solid Finite Element Model for Fatigue Analyses of a PWR Reactor Coolant System
Oriol Costa Garrido, Leon Cizelj, Igor Simonovski Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
The extension of operational license of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of
54 Nuclear Energy f0r New Eur0pe 2010 affected components. Analyses, based upon the in-service transient loads should be compared to the design-based ones.
The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements. The finite elements density is crucial for both the accuracy and the cost of the analysis.
The main goal of the paper is to propose a set of computational tools which assist the user in the deployment of modular spatial finite elements of the main components of the reactor coolant system, e.g. pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in an assembly. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. All components are meshed with hexahedral elements with quadratic interpolation.
The numerical examples provided in the paper include a parametric study quantifying the possible range of fatigue loads caused by a selected operational transient. This is followed by a study illustrating the influence of the mesh density in assessing the temperature fields of a fast thermal transient.
_605 ______
Some Consistency and Quality Tests for Finite Element Models of Grain and Grain Boundaries in Polycrystals
Mihaela Irina Uplaznik, Leon Cizelj, Igor Simonovski Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected] , [email protected], [email protected]
Of particular importance regarding degradation of components in nuclear power plants is the prediction of material ageing in view of development of intergranular damage. A variety of approaches, which tried to predict the effective overall behaviour of the polycrystalline aggregate from a known behaviour of the monocrystal, were developed over the years. Recent fast development of computers enabled development of polycrystalline aggregate models to study the effects of the microstructure on the load carrying capabilities of materials. Thus, Barbe et al. [1] and Cailletaud et al. [2] introduced a spatial model which uses stochastic methods such as Voronoi tessellation to accommodate the grain structure. A spatial model of a polycrystalline aggregate that accounts for random grain sizes, shapes and orientations is proposed here. Finite element method is used to obtain numerical solutions of the strain and stress fields. Qualitative (empirical distributions) and quantitative tests (Kolmogorov-Smirnov test) of the computed stress/strain tensors at all integration points will show the effects of different finite element types and mesh densities on the polycrystalline aggregate model. Furthermore, the mesh quality at the grain boundaries will be shown comparing the calculated normal stresses in all integration points of all cohesive elements with their theoretical expected values.
[1] F. Barbe, L. Decker, D. Jeulin and G. Cailletaud, Int.J. of Plasticity, 17, 513-536 (2001)
[2] G. Cailletaud, S. Forest, D. Jeulin, F. Feyel, I. Galliet, V. Mounoury and S. Quilici, Comp.Matr. Science 27, 351-374 (2003)
55 Nuclear Energy f0r New Eur0pe 2010
_606 ______
Damage Domain Approach as a Strategy of Damage Exceedance Frequency Computation.
José M. Izquierdo, Javier Hortal, Miguel Sánchez, Enrique Meléndez Consejo de Seguridad Nuclear, c/ Justo Dorado 11, 28040 Madrid, Spain [email protected], [email protected], [email protected], [email protected]
César Queral, Luisa Ibáñez, Israel Cañamón, Ernesto Villalba ETSI Minas (Universidad Politécnica de Madrid), c/ Alenza 4, 28003 Madrid, Spain [email protected], [email protected], [email protected], [email protected]
Jesús Gil, Iván Fernández, Santiago Murcia, J. J. Gómez Indizen Technologies S.L., av/ Pablo Iglesias 2-3ºB-2, 28003 Madrid, Spain [email protected], [email protected], [email protected], [email protected]
The process of Safety Margin Assessment, as proposed in the SMAP framework (Safety Margins Action Plan), is based on the extensive application of uncertainty analysis methods with the objective of obtaining an estimation of the exceedance frequencies of specified limits. The Damage Domain approach is presented as an adequate method to perform the uncertainty analysis, especially suited for those sequences where some events occur at uncertain times. This approach has been proposed by Consejo de Seguridad Nuclear (CSN), Spanish Nuclear Regulatory Body, in the context of the application exercise SM2A (Safety Margin Application and Assessment) and has been used to develop the analysis of selected scenarios of loss of Component Cooling and/or Service Water as a contribution of CSN to SM2A. Computations have been performed with the integral code MAAP and results are being substantiated with the best estimate plant simulation code TRACE. SCAIS (Simulation Code System for Integrated Safety Assessment) is the simulation-based computational framework that is being developed by the CSN to perform those computations.
_607 ______
Slovenian Regulatory Approach to Design Lifetime Extension
Artur Mühleisen, Matjaž Podjavoršek, Andreja Peršič, Djordje Vojnovič, Andrej Stritar Slovenian Nuclear Safety Administration, Železna cesta 16, SI-1000 Ljubljana, Slovenia e-mail: [email protected]
Nowadays many NPPs plan to operate longer than foreseen by their original design. Slovenia has a single NPP, Krško NPP that has been designed for 40 years of operation, which expires in 2023. It has already expressed the intention to operate for at least 60 years. Slovenian Nuclear Safety Administration (SNSA) had to form relevant regulatory requirements and its approach to approval of a prolonged operation. The requirements and the approach here presented have been formed conservatively, based on best international practices and specific Slovenian circumstances.
Circumstances, that substantially affected SNSA’s approach, are the national regulatory framework with the Periodic Safety Review (PSR) as a tool to periodically prove NPP’s nuclear safety, small size of the regulatory body, emerging NPP’s Aging Management Programme, U.S. origin of the NPP and European harmonization efforts. National regulatory framework sets PSRs at 10 years intervals as a
56 Nuclear Energy f0r New Eur0pe 2010 condition for extension of operating licenses. Therefore for the time being SNSA could only approve the design lifetime extension while the approval of prolonged operation has to wait for the approval of subsequent PSRs in 2013 and 2023.
Krško NPP submitted its incomplete request for approval of a prolonged design lifetime including relevant final safety analysis report changes at the end of March 2009. SNSA requested an supplement with an independent expert opinion with the scope equivalent to the scope of US NRC Safety Evaluation Reports prepared as a part of US NRC review of US license renewal applications. SNSA expects to issue a decision on NPP Krško design lifetime extension within this year.
_608 ______
High-effective Control System for Reactor Technological Equipment
V.I. Surin, N.A. Evstyukhin, Yu.A.Kapralov, A.A.Morozov National Research Nuclear University, Moscow, Russia E-mail: [email protected]
Diagnostics of reactor equipment which is in operation for a long time, detects that process of dangerous macroscopic defects formation, such as, circular or longitudinal cracks, has phase nature.
Distribution of cracks’ bedding by sizes or depth in power elements of construction (brackets, framework, bases, bearings and etc.) it is possible to receive, using of non-destructive methods of control.
The modern non-destructive methods of control (x-ray, ultrasonic, eddy-current) allow to diagnose confidently volume cracks with the sizes from the several tenth millimeters on depth of several millimeters bedding.
The electrophysical control gives the information on initial stages of origin and development of cracks in operating conditions.
The advantage of the given kind of the control consists that the gauge (or some gauges), having the linear sizes of an order of several millimeters, can be placed in that part of the process equipment, where formation of cracks most possibly. Diminutiveness and high reliability of gauges’ work allow to conduct continuous monitoring behind object of the control in inaccessible places. The particular interest represents use of the electrophysical control for diagnostics of welded seams in butt joint in the conditions of action of the corrosion environment, and also in working premises or halls with the raised radiating background.
Developed information-analytical monitoring systems (IAMS) should meet requirements of the state, branch and international standards.
By working out IAMS for reactor process equipment it is necessary to consider following main requirements: • working temperatures and a working environment; • value of mechanical loadings (pressure) in elements of constructions; • a prospective interval of change of electrophysical characteristics in the set temperature-power interval; • duration of diagnosing; • registration accuracy of the control parameters in the set conditions and an error of measurements; • constants time of elements operation of measuring system.
57 Nuclear Energy f0r New Eur0pe 2010
After the analysis of service conditions of the equipment it is possible to make a well-founded choice of block diagram IAMS, to define parameters of digitization of measured record characteristics and the requirement to a design of the measuring device or knot and other requirements.
_609 ______
Analysis of Cracks Resulting from Thermite Welding of Cathodic Protection
Marjan Suban, Simon Božič, Andrej Zajec, Robert Cvelbar, Borut Bundara Institute of Metal Constructions Mencingerjeva 7, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected], [email protected], [email protected]
In various steel pipes that are exposed to corrosion, they are protected with the cathodic protection where thermite welding of a copper conductor on a steel pipe is used. During the welding process and due to the nature of it, steel in the solid state comes in a contact with liquid copper. Contact of steel with liquid metal in some cases cause phenomenon known as the liquid-metal embrittlement or LME. Phenomenon was previously studied in cases such as soldering, but for thermite welding no records were found in accessible literature. The purpose of this paper is to draw attention to some irregularities and consequences arising from it, which in this type of welding can occur. At the end of article some measures to reduce these risks are given.
_610 ______
AREVA’s Fatigue Concept (AFC) – an Integrated and Multidisciplinary Approach to the Fatigue Assessment of NPP Components
Christian Poeckl, Steffen Bergholz, Dr. Juergen Rudolph, Nikolaus Wirtz AREVA NP GmbH Paul-Gossen-Strasse 100, 91052 Erlangen, Germany [email protected] , [email protected], [email protected], [email protected]
The prevention of fatigue damages is to be considered as a crucial issue in the light of changing boundary conditions: modification of the code based approaches, lifetime extension, new plants with scheduled operating periods of 60 years (e.g. EPR, KERENA) and improvement of disposability. The AREVA fatigue concept provides for a multiple step and multidisciplinary process (process engineering, fatigue monitoring, fatigue analyses etc.) against fatigue before and during the entire operation of nuclear power plants. Fatigue analyses are based on the real operational loads measured continuously on site in the plant. The entire process of fatigue design is based on an installed fatigue monitoring system. Qualified fatigue usage factors can be determined for the whole lifetime of the plant. Locations of potential fatigue failure are reliably identified and all efforts can be concentrated on these fatigue critical components.
The direct processing of the measured temperatures is immediately used for a first fast fatigue evaluation. This procedure is highly automated and allows for a rough evaluation of the recent usage factor as well as the qualitative comparability of the data.
58 Nuclear Energy f0r New Eur0pe 2010
In the framework of the Periodic Safety Inspection (PSI) a detailed fatigue check conforming to the code rules is carried out in order to determine the current state of the plant. This fatigue check is based on the real loads (specification of thermal transient loads) and finite element analyses in connection with the local strain approach to design against fatigue. The finite element analyses always include transient thermal determination of the temperature field and subsequent determination of (local) stresses and strains. The latter analyses might be simplified elastic or fully elastic plastic.
One peculiarity is the additional check against progressive plastic deformation (ratcheting) which is demanded by the design code. In the case of the elastic plastic approach much care has to be taken with respect to the application of an appropriate material law. Advanced nonlinear kinematic material laws of the Ohno & Wang type are favoured at AREVA at the present time. The implementation of these material laws within commercial finite element codes is still to be considered as non-standard use (especially for non-isothermal conditions) and an additional problem of experimental parameter identification and optimization is present.
As a conclusion, one essential benefit for the customer to apply the AREVA fatigue concept can easily be identified. Locations of potential fatigue failure are reliably identified and all efforts can be concentrated on these fatigue critical components. Thus, expensive costs for inspection can be essentially reduced. Of course, one presumption is the application of AREVA temperature measurement system FAMOS or similar in the power plant.
In the future, it is planned to integrate more sophisticated analyses concepts for fatigue damage (application of short crack fracture mechanics to fatigue crack growth) and realistic ratcheting simulations in the integral AREVA concept.
_611 ______
The Electronic Structure and Electrophysical Properties of Perspective Nuclear Fuel
V.I. Surin, N.A. Evstyukhin, Yu.A.Kapralov, E.Yu.Kapralov National Research Nuclear University, Moscow, Russia E-mail: [email protected]
The further progress in atomic energy is inconceivable without usage of Fast Breeder Reactors working in closed fuel cycle. For the new generation of Fast Breeder Reactors is supposed to use not only plutonium oxide and uranium (MOX fuel), but also the nitride fuel.
The carbide and nitride fuel is considered among perspective as has higher indicators of nuclear density and heat conductivity in comparison with traditional oxide fuel. Research of electrophysical properties of fuel under under irradiation (electric resistance, termo-emf, contact potential difference) allows to clear mechanisms of point defects formation and their evolution, mechanism of swelling, radiation forming and other radiating effects.
Unfortunately, in the literature practically there are no results of basic researches of electronic structure communication of fuel connections with physical properties. In the report was made an attempt at generalization of research’s results of uranium electronic structure connections with a crystal lattice of type NaCl, and also explanations of in-pile electrical conductivity of these connections on the basis of the analysis of electronic states density. For the purpose of details revealing of an electronic structure and character of dispersive curves in calculations used the adapted Slater-Koster method of a linear combination of atomic orbital.
59 Nuclear Energy f0r New Eur0pe 2010
Probabilistic Safety Assessment
60 Nuclear Energy f0r New Eur0pe 2010
_701 ______
Sensitivity Analysis of Emergency Diesel Generator Availability in IR-360 Nuclear Power Plant Using Fault Tree Method
Faramarz Yousefpour, Davood Babazadeh, Hamid Soltani Management of Nuclear Power Plant Company (MASNA), P.O. Box 14395-1359, Tehran,Iran Email address: [email protected]
Several studies demonstrate that core damage frequency due to Station Blackout (SBO) is highly sensitive to emergency diesel generator (EDG) performance. It is obvious that low emergency diesel generator performance and high quantity of diesel generator unavailability can significantly increase this SBO risk. Based on concerns about core damage frequency and associated reliability of emergency diesel generators, it is recommended to evaluate availability of EDGs in new nuclear power plants using Probabilistic Safety Assessment (PSA). Fault-tree analysis is a nominated approach which can be applied to systems in order to determine the probability of system’s failure modes. When the probabilities of systems unavailability are deemed not acceptable in fault tree analysis, the Importance and Sensitivity Analysis is applied to improve the system safety through design modifications. The Importance and Sensitivity Analysis is a consolidated procedure applied during the system’s design phase to identify the weakest parts of the system, i.e. those components whose failure modes give the greatest contribution to the likelihood of occurrence of the top event. Once these components are identified, suitable design modifications can be considered aiming at reducing the system failure probability. In order to obviate risk concerns, a comprehensive sensitivity analysis for applied fault tree of EGDs has been implemented by Safety Department of MASNA Company to find the weak points of EDG design in under design IR-360 nuclear power plant. This sensitivity analysis shows that how EDG unavailability in IR-360 depends on the failures of supporting systems such as cooling components. By means of the sensitivity analysis results, the proposed modifications in supporting systems definitely help to reduce final core damage frequency in IR-360 Nuclear Power Plant.
_702 ______
Sensitivity and Uncertainty Analysis for Age-Dependent Model of Test and Maintenance
Duško Kančev “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
Marko Čepin Faculty of Electrical Engineering, University of Ljubljana Tržaska 25, SI-1000 Ljubljana, Slovenia [email protected]
One of the principal activities for applications of risk-informed regulatory processes in the nuclear field is the ranking of structures, systems and components with respect to their safety-significance.
61 Nuclear Energy f0r New Eur0pe 2010
With reference to safety systems in particular, this requires an analysis of how the safety performance is affected by the stochastic behaviour of the components constituting the system. Different importance measures and various sensitivity indexes are often used for this scope.
On the other side, the large uncertainties of the ageing parameters as well as the uncertainties associated with the most of the reliability data collections are widely acknowledged. Thus, when realistic issues of system operation are included in the process of deriving the optimal scheduling of test and maintenance activities, such as the component ageing along with its parametric uncertainty, the computation of the mentioned importance measures and sensitivity indexes is not straightforward.
This paper deals with sensitivity and uncertainty analysis conducted on an age-dependent model of the test and maintenance activities for a selected stand-by safety system in nuclear power plant. The most important problem is the lack of data concerning the effects of ageing, which would correspond to more detailed modelling of ageing.
The obtained results indicate the extent that the uncertainty of the available ageing data sets has on the performed calculations.
_703 ______
Development of Probabilistic Seismic Hazard Analysis for Various International Locations; Challenges and Guidelines
Dr. Antonio Fernandez, P.E., Dr. Paul C. Rizzo, P.E. Paul C. Rizzo Associates, Inc. 500 Penn Center Blvd., Penn Center East, Suite 100 Pittsburgh, Pennsylvania, U.S.A 15235 [email protected], [email protected]
The study of seismic hazard is, without a doubt, a topic of significant interest for the analysis, design, and construction of engineering infrastructure. Conventional structures, such as private building facilities, small to medium bridges, and small to medium public works are commonly analyzed, designed and built with the aid of building codes and regulations. In general, building codes are applicable to regions where sufficient earthquake data is available, extensive research is well documented, and the performance criteria of the code in place are applicable and consistent with the expected lifespan and functionality of the structure. These conditions are typical of non-developed, low population areas for which not enough data and research are available, and therefore a site specific seismic hazard analysis is required.
This article provides guidance to conduct a site specific seismic hazard study, giving suggestions for overcoming those challenges that are inherent to the significant amounts of epistemic uncertainty for sites at remote locations. The text follows the general process of a seismic hazard study, describing both the deterministic and probabilistic approaches. Key and controversial items are identified in the areas of recorded seismicity, seismic sources, magnitude, ground motion models, and local site effects. A case history corresponding to a seismic hazard study in the Middle East for a Greenfield site in a remote location is incorporated along the development of the recommendations. Other examples of analysis case histories throughout the World are presented as well.
62 Nuclear Energy f0r New Eur0pe 2010
_704 ______
Optimal Generation Schedule of Power System Considering Nuclear Power Plant with Application of Genetic Algorithms
Blaže Gjorgiev a, Marko Čepin b, Atanas Iliev c a Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected]
b Faculty of Electrical Engineering, University of Ljubljana Trzaska 25, SI-1000 Ljubljana, Slovenia [email protected]
c Faculty of Electrical Engineering and Information Technologies – Skopje, Karpoš II bb, 1000 Skopje, Macedonia [email protected]
Power system generation scheduling is an important issue from technical, economical and environmental safety viewpoints. It involves multiple, sometimes conflicting optimization criteria for which no single optimal solution with respect to all criteria can be determined. Nuclear safety is one of the key parameters for power systems, which include nuclear power plants.
Method for optimal generation schedule of power plants in electrical power system with genetic algorithms is developed. Maximization of high level of nuclear safety, minimization of costs and minimization of the emission of pollutants from thermal power plants are emphasized among the goals of the results.
The system constraints are modeled in a way that the behavior of the power system model matches as close as possible to the real conditions of operation. Thus, constraints are solved using fuzzy- penalization. Main feature of fuzzy-penalization is that the method use fuzzy sets of crisp values for each constraint. These sets are identified with various types of fuzzy functions, which are constructed for the purpose of specific constraint.
The feature of new generation of nuclear power plants, which can follow the load, is included in the constraints.
The expected results indicate how the nuclear power plant should operate and how the other power plants in the system should operate in order that the overall production costs and emission of pollutants are minimized.
63 Nuclear Energy f0r New Eur0pe 2010
_705 ______
Advantages and Difficulties with the Application of Methods of Probabilistic Safety Assessment to the Power Systems Reliability
Marko Čepin Faculty of Electrical Engineering, University of Ljubljana Tržaska 25, SI-1000 Ljubljana, Slovenia [email protected]
Andrija Volkanovski “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
The probabilistic safety assessment is a standardized and widely accepted method for assessing and improving the reliability and safety of complex technologies, which is particularly true for the nuclear and the aerospace industry. The number of its applications to reliability of power systems is increasing.
The objective of the paper is to collect and present the advantages and difficulties with the application of methods of probabilistic safety assessment to power systems reliability. The methods which apply the features of fault tree analysis to reliability analysis of power systems are presented and compared.
The results are discussed and evaluated. The main advantages are connected with the facts that the methods are relative simple, the models are mostly available and that the data is intensively collected, evaluated and can serve as an excellent support for the quantitative part of the analyses.
The difficulties are connected with different nature of components interaction within the power system compared to the safety systems in a nuclear power plant and with the large complexity of the power system, which distinguishes the term reliability into its static and dynamic parts known as adequacy and security.
_706 ______
Ageing Prioritization Based on PSA Results
Andrija Volkanovski “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
Method for prioritization of the nuclear power plant components due to ageing based on the results of probabilistic safety assessment is developed and presented. Ageing is a process, where the properties of systems and processes may degrade through the time and age. The prioritization of the components is done with new importance coefficient obtained from the Fussel-Vesely importance measure of the component and utilization of Taylor expansion approach. The change of the core damage frequency for different replacement and surveillance intervals is calculated from the new importance coefficient. Components are sorted depending on their contribution to the change of the core damage frequency resulting from their ageing.
64 Nuclear Energy f0r New Eur0pe 2010
Developed method is applied on the probabilistic safety assessment model of the test nuclear power plant. The TIRGALEX database was used as input for the components ageing rates. The obtained results include the change of the core damage frequency from the individual components for different test and replacement intervals. The overall change of the core damage frequency is calculated and presented.
The limitations of the developed method and possibilities for future improvement are discussed. The major issues and problems considering the introduction of ageing in the probabilistic safety assessment are presented.
65 Nuclear Energy f0r New Eur0pe 2010
Severe Accidents
66 Nuclear Energy f0r New Eur0pe 2010
_801 ______
Implementation of Severe Accident Management Guidelines to Shutdown and Low-Power Operation Modes for VVER and PWR Plants
Oleg Solovjanov [email protected] Robert Lutz Nathalie Dessars Thibaut Rensonnet Westinghouse Electric Belgium, S.A. Rue de L’Industrie 43 1400 Nivelles, Belgium
About 15 years ago Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) were developed to offer the Technical Support Center guidance to mitigate accidents involving core damage. Since then, significant improvements have been identified through plant specific SAMG validations and implementations which resulted in a new PWROG SAMG (Revision 1) issued in 2001.
The primary goals of the WOG SAMG are to terminate fission product releases from the plant, to prevent failure of intact fission product boundaries and to return the plant to a controlled stable state.
The Generic WOG SAMGs are designed to mitigate low probability core damage events initiated while the plant was at power. However, recent Probabilistic Risk Analyses (PRA) have been extended to shutdown and low power operation modes in many countries. Many studies such as the shutdown PRA for Beznau, Koeberg, EdF 900/1300, and VVER plants in Central Europe (Hungary, Slovak and Czech Republic) as well as latest industry events, such as Paks NPP shutdown fuel damage accident, demonstrated that the core damage frequency from an accident occurring when at shutdown or low power operation modes was of the same order of magnitude and even higher (up to 80% of CDF for some plants) than the one at power.
As a result, it was concluded that Accident Management Measures (EOP and SAMG) should be extended to cover the full range of possible plant operation modes. Westinghouse has developed a methodology to extend the applicability of the WOG SAM Guidelines to shutdown and low power conditions. The principal changes required in the entry conditions, diagnostic parameters, diagnostic prioritization, as well as specific severe accident guidelines were identified and implemented.
The SSAMG package is integrated into at-power SAMG to form a complete symptom-based SAMG package applicable to all Plant Operational States (POS). Also, accidents occurring when the fuel is not in the core (including the spent fuel pool) were considered.
Shutdown SAMG (SSAMG) based on this approach has been implemented at PWR plants (Koeberg and Beznau NPPs) and VVER-440 plants (Paks and Mochovce 3&4 NPPs).
The purpose of this paper is to present how the WOG SAMG has been adapted to cover all POSs. The key impact on the WOG SAMG package (entry conditions, diagnostic tools, control room and Technical Support Center guidelines, computational aids) will be discussed and the lessons learned from the implementation will be provided.
67 Nuclear Energy f0r New Eur0pe 2010
_802 ______
In-vessel Severe Accident Analysis of PWR Station Blackout with the RELAP5/SCDAP and ASTEC Codes
Milan Amižić, Siniša Šadek, Nenad Debrecin Faculty of Electrical Engineering and Computing, University of Zagreb Unska 3, 10000 Zagreb, Croatia [email protected], [email protected], [email protected]
The RELAP5/SCDAP is the most detailed and complete mechanistic code designed to predict behaviour of reactor systems during normal and accident conditions. The code consists of the SCDAP part, which deals with description of core heat-up and degradation in case of a severe accident, the COUPLE module which allows the simulation of a debris bed or molten pool in the lower head and the RELAP5 thermohydraulic code for a description of the behaviour of the reactor coolant system and other required systems.
The ASTEC is fast running integral code, developed by the IRSN, France, and GRS, Germany, which allows the calculation of the entire sequence of a severe accident in water-cooled reactors, covering all important in-vessel and ex-vessel phenomena.
In this paper in-vessel severe accident phenomena caused by station blackout in the NPP Krško are presented. It was assumed that both off-site and on-site AC power are unavailable, therefore, primary system coolant inventory was decreasing due to water leakage from the reactor coolant system through reactor coolant pump seals. Hence, water was injected into reactor coolant system only from accumulators.
The objectives of the paper are to present the preliminary model of NPP Krško for the ASTEC calculation and to compare obtained results with the RELAP5/SCDAP results.
_803 ______
LACOMECO Experimental Platform at KIT
A. Miassoedov, X. Gaus-Liu, T. Jordan, L. Meyer, W. Tromm, M. Steinbrück, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen, Germany [email protected], [email protected]
The platform LACOMECO at Karlsruhe Institute of Technology (KIT) provides to European research institutions access to several experimental facilities which are designed to study the remaining severe accident safety issues, including the coolability of a degraded core, corium coolability in the RPV, possible melt dispersion to the reactor cavity and hydrogen mixing and combustion in the containment. These facilities are unique in providing experimental programmes in specific fields of core damage initiation up to hydrogen behaviour and are designed to be complementary to other European facilities and experimental platforms to form a coherent European nuclear experimental network.
The specific research of the addressed phenomena involves very substantial human and financial resources and, in general, the research field is too wide to allow investigation of all phenomena by any national programme. To optimise the use of the resources, the collaboration between nuclear utilities, industry groups, research centres and safety authorities, at both national and European levels is very
68 Nuclear Energy f0r New Eur0pe 2010 important. This is precisely the main objective of the LACOMECO platform, which aims to provide these resources and to facilitate this collaboration by offering the experimental platforms for the transnational access in Europe. LACOMECO project is strongly coupled with other European projects, such as SARNET2, as well as with third countries (Russian Federation, Ukraine, Kazakhstan) through the International Science and Technology Center (ISTC) and the Science & Technology Center in Ukraine (STCU). The experimental results are used for the development of models and their implementation in the severe accident codes such as ASTEC, thus preserving and diffusing this knowledge to a large number of current and future end-users throughout Europe.
The LACOMECO experimental platform at KIT includes:
- QUENCH facility is the only operating experimental facility within the European Union for investigations of the early and late phases of core degradation in prototypic geometry for different reactor designs and different cladding alloys, incl. analysis of the relocation of cladding and fuel and the formation and cooling of in-core debris beds to gain information on the characteristics of the created particles.
- LIVE facility concentrates on the investigation of the whole evolution of the in-vessel late phase of a severe accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting in large scale 3D geometry with emphasis on the transient behaviour.
- HYKA experimental facilities are among the largest available in the world. In combination with the high static and dynamic pressures the experimental facilities are designed for, a unique experimental centre especially for combustion experiments in confined spaces is available with HYKA. Due to the different orientations and sizes the set of large and strong experimental vessels offers a flexible basis for scientific experimental work on reactive hydrogen mixtures.
- DISCO is the only operating facility available worldwide for integral DCH investigations. It is designed to perform scaled experiments that simulate melt ejection from the RPV to the reactor cavity after the RPV failure under low system pressure during severe accidents in LWRs. These experiments investigate the fluid-dynamic, thermal and chemical processes during melt ejection out of a breach in the lower head of an LWR pressure vessel at pressures below 2 MPa.
69 Nuclear Energy f0r New Eur0pe 2010
_804 ______
Study of Boron Behaviour in the Primary Circuit of Water Reactors under Severe Accident Conditions; a Comparison of Recent Integral and Separate-Effect Data
Tim Haste, Frédéric Payot, Cristina Dominguez, Philippe March, Béatrice Simondi-Teisseire Institut de Radioprotection et de Sûreté Nucléaire (IRSN) Centre d'Etudes de Cadarache, BP 3 - 13115 Saint-Paul-Lez-Durance Cedex, France [email protected], [email protected], [email protected], [email protected], [email protected]
Martin Steinbrück Karlsruhe Institute for Technology Postfach 1, D-76021 Karlsruhe, Germany [email protected]
Roland Zeyen European Commission Joint Research Centre, Institute for Energy Centre d'Etudes de Cadarache, BP 3 - 13115 Saint-Paul-Lez-Durance Cedex, France [email protected]
Boron carbide (B4C) is widely used as an absorber material in many commercial reactors, such as boiling water reactors, Russian VVERs, later French pressurised water reactors, and in the European
Pressurised Water Reactor. Under hypothetical severe accident conditions, B4C reacts with its surrounding stainless steel cladding, producing eutectic melts above 1200°C; remaining bare B4C and B4C/metal mixtures are then exposed to steam and oxidize highly exothermically. As well as hydrogen, gases and aerosols containing boron and carbon compounds are produced, which affect the transport and deposition of radiologically important fission products such as iodine and caesium in the circuit, and subsequent behaviour in the containment.
The influence of a B4C control rod on fuel degradation and fission product release through to the late phase, material transport in the circuit and behaviour in the containment was studied in the nuclear- heated integral experiment Phebus FPT3 at Cadarache. Carbonaceous gas production was monitored, and evidence for substantial deposition of B-containing compounds in the circuit was gathered. The electrically-heated QUENCH-07 and QUENCH-09 tests at Karlsruhe used similar bundle geometry and had similar but less severe test conditions, and included a water reflood phase. Extensive separate- effects tests on oxidation of B4C and its interaction with surrounding materials have also been performed at IRSN Cadarache (BECARRE programme) and at Karlsruhe (BOX, LAVA, QUENCH- SR).
This paper compares the phenomena involving B4C observed in these experiments, concentrating on degradation, transport and deposition behaviour. In particular, there is evidence concerning blockage formation in the hot leg of the FPT3 circuit, which affects the transport of fission products, and this is supported by similar observations in the separate-effect tests. Possible explanations for the phenomena observed are advanced, and suggestions for further work to improve understanding are put forward.
70 Nuclear Energy f0r New Eur0pe 2010
_805 ______
Current Activities Of CSN in the Field of Severe Accident Research
Fernando Robledo Consejo de Seguridad Nuclear (CSN) CL. Pedro Justo Dorado Dellsmans, 11, 28040 Madrid- Spain [email protected]
Luis E. Herranz CIEMAT Av. Complutense, 22 28040 Madrid, Spain [email protected]
The cornerstone of the CSN activities in the field of severe accident research is the CSN-CIEMAT Technical Agreement in the field of severe accidents. The current main activities carried out under this Agreement are: 1. Participation in the ARTIST-II project. 2. Participation in the Sandia Fuel Project (SFP) of NEA/OECD. 3. Updating on outcomes from source term research. 4. To review and update the input decks for the analytical tools used by the CSN during emergency responses. 5. Updating of MELCOR models used by CSN in their assessments of PSA-2 studies and severe accident analyses. 6. Uncertainties analyses in severe accidents. Other CSN’s severe accidents research activities aside of this Agreement that are currently being performed are: 7. Participation in the MCCI-2 project. 8. Application of CSN developed methods and tools of Dynamic Reliability to safety assessment on H2 combustion. Main activities and results performed within this research program will be reported in this paper.
71 Nuclear Energy f0r New Eur0pe 2010
_806 ______
Developing New Methodology for Nuclear Vulnerability Assessment
Venceslav Kostadinov Slovenian Nuclear Safety Administration Železna cesta 16, SI-1000 Ljubljana, Slovenia [email protected]
The fundamental aim of an efficient emergency preparedness and response system is it to provide sustained emergency readiness and to prevent emergency situations and accidents. But when an event occurs, the mission is to mitigate consequences and to protect people and the environment against nuclear and radiological damage. The emergency response system, which would be activated in the case of a nuclear and/or radiological emergency and release of radioactivity to the environment, is an important element of a comprehensive national protection system of nuclear and radiation safety.
In the past, national emergency systems explicitly did not include vulnerability assessments of the critical nuclear infrastructure as an important part of a comprehensive preparedness framework. But after the huge terrorist attack on 11.09.2001, decision-makers became aware that critical nuclear infrastructure could also be an attractive target to terrorism, with the purpose of using the physical and radioactive properties of the nuclear material to cause mass casualties, property damage, and detrimental economic and/or environmental impacts. The necessity to evaluate critical nuclear infrastructure vulnerability to threats like human errors, terrorist attacks and natural disasters, as well as preparation of emergency response plans with estimation of optimized costs, are of vital importance for assurance of safe nuclear facilities operation and national security.
In this paper presented new methodology and solution methods for vulnerability assessment can help the overall national energy sector to identify and understand the terrorist threats to and vulnerabilities of its critical infrastructure. Moreover, adopted methodology could help national agencies to develop and implement a vulnerability awareness and education programs for their critical assets to enhance the security and a safe operation of the entire energy infrastructure. New methods can also assist nuclear power plants to develop, validate, and disseminate assessment and surveys of new efficient countermeasures. Consequently, concise description of developed new quantitative method and adapted new methodology for nuclear vulnerability assessment of nuclear power plants are presented.
_807 ______
Analysis of Melt Droplets Crust Growth During Steam Explosion Premixing Phase
Matjaž Leskovar, Mitja Uršič Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
A steam explosion may occur when, during a severe nuclear reactor accident, the molten core comes into contact with the coolant water. In this energetic fuel coolant interaction process the energy of the molten corium is transferred to the coolant water in a timescale smaller than the timescale for pressure relief. This may lead to the formation of shock waves and production of missiles that may endanger
72 Nuclear Energy f0r New Eur0pe 2010 surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to the direct release of radioactive material to the environment.
The strength of a steam explosion depends on the mass of melt droplets, which can efficiently participate in the steam explosion – that is the mass of liquid melt droplets in regions with high water content. The mass of liquid melt droplets is defined by the competing processes of droplets generation during jet breakup and droplets solidification during the heat transfer to the surrounding water. Therefore the melt solidification is one of the most decisive processes influencing the steam explosion energetics.
In the paper the melt droplets solidification during the premixing phase will be analyzed and discussed. A melt droplet crust growth model, which treats the melt droplet as an opaque symmetrical sphere, was developed. The model assumes that the heat inside the droplet is transferred only by conduction and that the heat losses from the droplets surface are uniform. The heat flux from the droplets surface is determined as the sum of the film boiling heat flux and the radiation heat flux. The temperature profile inside the melt droplet is calculated numerically with the finite differences method. With the model, the influence of the droplets size, melting temperature, melt superheat, thermal properties and ambient conditions on the droplets crust growth was investigated and the most influential parameters were determined. Based on the performed study, some insight in the observed differences in the behaviour between simulant and prototypic melts will be given and the extrapolation of experimental findings to reactor conditions will be discussed.
_808 ______
Peer Review of Trial Application of Low Power and Shutdown PRA Standard
Srinivasa Visweswaran, Program Manager, Risk Applications and Methods II, Westinghouse Electric Company, [email protected]
David Finnicum, Fellow Engineer, Risk Applications and Methods II, Westinghouse Electric Company, [email protected]
Purpose of this paper is to present the lessons-learned from the peer review at a non-US reactor site of the LPSD PRA performed against the draft Low Power and Shutdown (LPSD) Standard developed by the American Nuclear Society (ANS).
In 1995, the U.S. Nuclear Regulatory Commission (USNRC) issued a Policy Statement (Ref. 1) on the use of probabilistic risk assessments (PRA), encouraging its use in all regulatory matters. Since that time, many uses have been implemented by the nuclear plan licensees. Consequently, confidence in the information derived from a PRA is an important issue. Through Regulatory Guide 1.200, Rev. 2, (Ref. 2), the USNRC has endorsed the Combined PRA Standard issued by ASME/ANS (Ref. 3) for developing internal and external event PRAs while at-power. This Standard has been used extensively by US nuclear plants for carrying out the at-power internal event PRAs, and more recently, for their Fire PRAs. Each of the at-power internal event and Fire PRAs has undergone a peer review, using a written process, per the requirements of the Standard.
Several US plants have developed an LPSD PRA without an LPSD PRA Standard available. The ANS has been working on an LPSD PRA Standard for several years. The most recent revision of the LPSD PRA Standard (Ref. 3), is currently being revised based on comments received during the balloting in 2009. The draft standard will soon be out for ballot again and after approval, it will be available for application. 73 Nuclear Energy f0r New Eur0pe 2010
One non-US plant, a 3-loop PWR design similar to that of North Anna, has completed a detailed at- power as well as LPSD PRA. The plant wanted their at-power PRA to be peer reviewed against the Combined Standard and the LPSD PRA to be peer reviewed against the draft ANS LPSD PRA Standard. Westinghouse Electric Company assembled a team of eight PRA experts from various parts of the world and conducted the peer review at the customer’s site during the week of April 11-15, 2010. Since a written process was not available for the peer review of LPSD PRA, Westinghouse used the written process developed for the at-power PRA and reviewed the at-power and LPSD portions of the PRA in parallel.
The scope of the LPSD PRA included evaluation of core damage frequency (CDF) and large early release frequency (LERF) for all the Plant Operating States (POS) associated with the low power and shutdown operation. A total of 19 POSs had been identified and the fraction of time the plant resided in each POS had been calculated. The Review Team decided to carry out the review in the same one- week period as is done individually for the internal event and fire PRAs by using additional review members (eight instead of six)
Since the Review Team was very experienced in conducting peer reviews, it was also able to assess the adequacy of the draft LPSD Standard. The Review Team noted that the technical elements of the LPSD Standard were generally similar to the ones in the at-power Combined Standard, except that fire PRA is not part of the scope. A new element, Plant Operating State, with 22 supporting requirements (SR) has been added to the LPSD Standard. For the remaining technical elements that were similar to that of the at-power Standard, 157 SRs remained unchanged, 41 SRs were expanded via notes, 56 were expanded and 23 new SRs were added.
The Review Team found that where it needed to spend extra time was on a new LPSD PRA technical element POS. For all other technical elements, much of the requirements were similar to that for the full power internal events PRA. Additional, LPSD-specific requirements were clear and easy for the PRA team to address and the Review Team to review.
The Review Team has identified a few specific requirements that were either not clear or specific enough. The Team had specific observations about the technical element on human reliability analysis (HRA). The Review Team concluded that the LPSD PRA Standard is well written and provides appropriate guidance to the PRA practitioners in developing the LPSD PRA.
References:
60 FR 42622, “Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy Statement,” Federal Register, Volume 60, Number 42622, August 16, 1995.
Regulatory Guide 1.200, Rev.2 “An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities”, March 2009
ANSI/ANS-58-22-20xx,” Low Power and Shutdown PRA Methodology”, approved November xx 2009
74 Nuclear Energy f0r New Eur0pe 2010
_809 ______
Modelling of Ruthenium Release in Air and Steam Atmospheres under Severe Accident Conditions using the MAAP4 Code
Emilie Beuzet, Jean-Sylvestre Lamy EDF R&D 1 Avenue du Général de Gaulle, F-92140 Clamart, France [email protected], [email protected]
Hadrien Perron EDF R&D Avenue des Renardières, Ecuelles, F-77818 Moret sur Loing, France [email protected]
Eric Simoni Institut de Physique Nucléaire Université Paris Sud XI, F-91406 Orsay, France [email protected]
In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and Fission Products (FPs) release in the environment (Source Term). Water vaporisation and core uncovery can occur due to residual power. These phenomena enhance core degradation and subsequently, molten materials can relocate in the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and an air ingress. After failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in core. Indeed, Zircaloy-4 cladding oxidation is very exothermal and fuel interaction with the cladding material decreases its melting temperature by several hundreds of Kelvin. FPs release can thus be increased, noticeably that of ruthenium. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as ruthenium gaseous tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues.
A review of existing data on these phenomena reaches a relatively well understanding. In terms of oxygen affinity, the fuel is oxidised before ruthenium, from UO2 to UO2+x. Its oxidation is a rate- controlling surface exchange reaction with the atmosphere, so that stœchiometric deviation and oxygen partial pressure increase. High temperatures combined with oxygen in the atmosphere lead to fuel expansion and cracks. In these conditions, intra-and inter-granular diffusions of ruthenium in the fuel matrix are so enhanced that it is possible to consider an instantaneous volatilisation of ruthenium oxides.
Based on these considerations, a completely new model has been implemented in the EDF local version of MAAP4.07 severe accident code (Modular Accident Analysis Program). The fuel oxidation modelling takes into account many kinds of atmospheres (steam and/or air and/or hydrogen), the stœchiometric evolution and the oxygen partial pressure of the fuel matrix. The release of ruthenium oxides is calculated considering their particular reaction constants. The model was assessed by the simulation of different VERCORS experiments in air, steam and mixed atmospheres. These experiments, conducted at CEA, are specially designed to study FPs release from fuel under different atmospheres. This paper deals with the main results obtained with MAAP4.07 when simulating these tests.
75 Nuclear Energy f0r New Eur0pe 2010
_810 ______
Simulations of KROTOS Alumina and Corium Experiments: Applicability of the Improved Solidification Influence Modelling
Mitja Uršič, Matjaž Leskovar, Borut Mavko Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
A steam explosion may result from the rapid and intense heat transfer that may follow the interaction between the molten material and coolant. KROTOS steam explosion experiments have revealed important differences in behaviour between simulant alumina and oxidic corium melts. Differences in the material physical properties are one of the probable reasons for the observed differences in the steam explosion efficiency. The experimentally observed differences are importantly attributed to the differences in the melt solidification and to the differences in the void production.
Due to the recognized importance of the solidification issue, an improved solidification influence model was developed and implemented into the Eulerian fuel-coolant interaction code MC3D, which is being developed by IRSN, France. The developed solidification influence model is based on an improved temperature profile modelling inside the melt droplet and the computation of the mechanical effect of the crust thickness on the fragmentation process. The purpose of the proposed solidification influence modelling was to improve the determination of the melt droplet mass, which can efficiently participate in the steam explosion process.
In the paper the general applicability of the implemented improved solidification influence modelling approach in the MC3D code will be examined. For that purpose the alumina K44 and corium K53 triggered steam explosion experiments, performed in the KROTOS facility, were simulated. The simulation results will be presented in comparison and discussed, focusing on the melt droplets solidification issue.
_811 ______
Numerical Investigation of Natural Circulation during a Small Break LOCA Scenarios in a PWR-System using the TRACE v5.0 code
E. Coscarelli, A. Del Nevo, F. D’Auria University of Pisa San Piero a Grado Nuclear Research Group (GRNSPG) Via Diotisalvi 2, 56122 [email protected], [email protected], [email protected]
The present paper deals with the analytical study of the PKL experiment F4.1 performed using the TRACE code (version 5.0). The test F4.1 executed in PKL-III facility investigates both, the boron dilution occurrence and the heat transfer from primary to secondary side under natural circulation conditions (single and two phase flow) and reflux condensation. The boron dilution events in condition of reduced mass inventory, namely during a SB-LOCA scenarios, are considered. The relevance of those transient is connected with the possibility that unborated coolant enters in the core, causing re-criticality or, worse, power excursions. The numerical investigation is performed by developed a complete TRACE input model of the PKL integral test facility, including secondary as well as primary system. The aim of this work is the assessment of the TRACE code against the boron
76 Nuclear Energy f0r New Eur0pe 2010 transport and the heat transfer mechanism in the different flow regimes that take place during the experiment. The accuracy of the calculation is evaluated by qualitative and quantitative analysis. The quantification of the accuracy is performed using the Fast Fourier Transform Base Method (FFTBM) developed at University of Pisa. The tool provides an integral representation of the accuracy quantification in the frequency domain.
_812 ______
Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V2 to Bottom Water Reflood Experiment PARAMETER-SF4
Vasiliev A.D. Nuclear Safety Institute (IBRAE) B.Tulskaya 52, 115191 Moscow, Russia [email protected]
The PARAMETER-SF4 test conditions simulated a severe LOCA (Loss of Coolant Accident) NPP (nuclear power plant) sequence in which the overheated up to 1700 - 2300K core would be reflooded from the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in July 21, 2009, and was the fourth of four experiments of series PARAMETER-SF.
PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents.
The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm- diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness.
After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF4 test, the bottom flooding was initiated.
The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF4 experiment.
The important feature of PARAMETER-SF4 test was the air ingress phase during which the air was supplied to the working section of experimental installation. It is known that zirconium oxidation in the air proceeds in different way in comparison to oxidation in the steam.
Thermal hydraulics in PARAMETER-SF4 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulic and chemical behaviour. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF4 test.
77 Nuclear Energy f0r New Eur0pe 2010
Radioactive Waste and Decommissioning
78 Nuclear Energy f0r New Eur0pe 2010
_901 ______
Perception of Radioactivity and Attitudes Towards Radioactive Waste
Nadja Železnik Agency for radwaste management Parmova 53, Ljubljana, Slovenia [email protected]
Siting of a radioactive waste repository, even for the waste of low and intermediate level (LILW) radioactivity, presents a great problem in almost every country that produces such waste. The main problem is not a technical one, but socio-psychological, namely the acceptability of this kind of repository. In 1996 the Slovenian national Agency for radwaste management - ARAO re-initiated the search for a low and intermediate level radioactive waste (LILW) repository location with a new, so- called combined approach to the site selection. In this context the influence of social models on acceptability was studied. Previous research on people’s perception of the LILW repository construction, their attitudes towards radioactive waste, their willingness to accept it, based on several surveys, indicated significant differences in answers of experts and lay persons, mainly regarding evaluation of the consequences of repository construction. This findings support the use of mental model approach.
Based on the findings of pilot investigations a mental model approach to the radioactivity, radioactive waste and repository was used as a method for development of risk communication strategies with local communities which volunteered to host the repository. The mental models were obtained by adjustment of the method developed by Morgan and co-workers (Risk Communication, 2002) where expert model of radioactivity is compared with mental model of lay people obtained through individual opened interviews. Additional information on trust and role of main actors in the site selection process was gained with the overall questionnaire on the representative sample of Slovenian population.
Results of the survey confirm some already known findings, in addition we gained new cognitions and with analyses obtained the relationships and ratios between different factors, which are characteristics both for the general public and for the public, which is involved in the site selection process for a longer period and has been living beside a nuclear power plant for one generation. People have in general negative associations regarding the repository, the perceived risk for nuclear facilities is high, and trust in representatives of governmental institutions is low. Mental models of radioactivity, radioactive waste and the LILW repository are mostly irregular and differ from the experts’ models. This is particularly valid for the models of radioactivity and the influences of radiation on people irrespective of the public which was involved in the survey. Among the most important factors which influence public acceptability of the construction of the LILW repository in the domestic location is perceived risk to the nuclear power plant. This factor is more important than knowledge on radioactivity and radioactive waste for different groups, also for the local public with experience of living beside nuclear power plant. Although it can be seen that the factor of knowledge has higher importance in the local community which means that communication activities among local citizens do influence the acceptability. Based on the analyses of the results, the starting points for improvement of communication plans were prepared, which should be used by the implementer of the site selection, and later during the repository construction. These communication starting points have a broader validity, since they could be suitable also for risk communications for other technologies.
79 Nuclear Energy f0r New Eur0pe 2010
_902 ______
Preliminary Design of Vrbina LILW Repository
Boštjan Duhovnik IBE, Consulting Engineers Hajdrihova 4, SI-1000 Ljubljana, Slovenia [email protected]
Janja Špiler Agency for Radwaste Management Parmova 53, SI-1000 Ljubljana, Slovenia [email protected]
Simultaneously with the siting of the repository, intensive project engineering activities relating to radwaste conditioning and disposal facility started at the end of 2004. In a Comparative Multilateral Study of Alternatives a solution of disposal in below-ground silos was proved to be the most suitable one for the Vrbina site. The solution is developed in detail in Preliminary Design documentation.
The planned LILW Repository will be located some hundred meters east of the Nuclear Power Plant Krško. The site was formally approved in the scope of the Decree on National Spatial Plan at the end of 2009. The repository is designed for disposal of half of short-lived LILW from NEK and of all the remaining Slovenian short-lived LILW (9,400 m3) with a possibility of further disposal capacities extension providing disposal of the overall LILW quantity (18,200 m3).
The repository will consist of a disposal part, composed of a set of modular disposal units (silos), and other structures required for acceptance, conditioning and storage of radioactive waste. Besides technological facilities, a visitor center, an administrative building and a service building are anticipated on the site as well. All buildings and disposal structures of the repository will be constructed on a platform which will protect them against floods.
Prior to disposal, all LILW will be inserted into concrete disposal containers of external dimensions 2.55 x 2.55 x 3.25 m. In an individual container, 9 tube-type containers (the most often type of package in Krško NPP) or 27 standard 200-liter drums or unpacked LILW with a volume of approximately 13 m3 will be placed. Voids in the container will be backfilled with backfilling mortar.
The containers will be disposed into a disposal silo of an inner diameter of 27.3 m and useful (net) height of 33 m. The bottom of the silo is approximately 55 m below the elevation of the handling platform. The LILW placed in the containers will be inserted into the silo by a portal gantry crane. After each of the ten layers with 70 disposal containers is filled-up, the remaining voids between the container walls and between the containers and the silo wall will be backfilled with backfilling material. Water which will eventually seep through the silo walls will be collected in the draining basis of the silo and drained via an inspection gallery to the access shaft, where sampling will be performed prior to its pumping to the surface.
80 Nuclear Energy f0r New Eur0pe 2010
_903 ______
Transport of Radioactive Waste from Small Producers
Marko Kostanjevec, Marija Fabjan, Simona Sučić ARAO – Agency for Radwaste Management Parmova 53, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
ARAO has been performing the public service of managing radioactive waste from medicine, industry and research institutions from the so called small producers and provides conditions for safe management of radioactive waste in line with the Governmental decree on the Mode, Subject and Terms of Performing the Public Service of Radioactive Waste Management (Official Gazette RS No. 32/99).
Besides other activities the ARAO provides also the transport of radioactive waste from producers' premises to Central storage facility in Brinje. ARAO is carried out all modes of transport of radioactive waste from small producers, including transport in case of emergency situation and in case of transport accidents or some other accidents.
Until the beginning of 2005, ARAO organized transport with the contractor. From April 2005, ARAO is qualified for the transport of radioactive waste. Parallel to the purchasing of the necessary equipment for the safe transport and tools for handling the waste, ARAO also prepared several procedures and working instruction for the operating of the transport activities. These include all operations and conditions of the design, preparation of packaging, consigning, carriage, transit and receipt at the final destination of the package of small producers, at the Central storage facility.
The transport of radioactive waste in Slovenia is regulated with the Transport of Dangerous Goods Act which was put into force in 2002 and ADR-European Agreement Concerning the International Carriage of Dangerous Goods by Road. ARAO has fulfilled all legislation requirements for transport. It has ADR drivers, a safety adviser, transport containers and equipped vehicle.
Various administrative and technical activities of the ARAO, which have been implemented in the frame of the transport of radioactive waste from small producers, are described in this paper.
_904 ______
Regulatory Experiences with the Unconditional Clearance of Radioactive Material in Slovenia
Polona Tavčar, Igor Osojnik, Maksimilijan Pečnik Slovenian Nuclear Safety Administration Železna cesta 16, p.p.5759, SI-1001 Ljubljana, Slovenia [email protected]
The unconditional clearance of material is defined as release of material from licensed radiation practices. After release the material is not subject of any further regulatory control. Such material shall contain activities of radionuclides below general clearance levels and can be released without any condition or restriction. General clearance levels in Slovenia were established in 2004 by the Decree on the radiation practices (Official Gazette No. 48/04 and 9/06). The decree also provides the general
81 Nuclear Energy f0r New Eur0pe 2010 criteria for conditional clearance of the material. This paper presents Slovenian experience only with unconditional clearance.
The concept of clearance in general is most widely used in the decommissioning and dismantling of nuclear installations. Currently there is no decommissioning of nuclear installations undergoing in Slovenia. Anyway during the renovations, conditioning of the radioactive waste or regular outages of the facilities some waste material is produced with no or negligible content of the radionuclides. These materials can be released unconditionally after notification of the regulatory body, if concentration of radionuclides is lower than general clearance levels.
The paper presents the regulatory experiences with the unconditional clearance of the material mostly coming from Krško Nuclear Power Plant and from Central Interim Storage Facility. The material unconditionally cleared from Krško Nuclear Power Plant was mostly scrap metal send for the recycling. After the clearance some other types of material such as spent ion exchange resins from secondary circuit and larger amount of charcoal from filters were sent to the collectors of specific dangerous waste.
The radioactive waste stored in Central Interim Storage Facility in Brinje was characterized and repacked recently. Through this process different material such as old metal drums, parts of the equipments, parts of dismantled ionization smoke detectors, some lead shielding and other material contaminated below general clearance levels, was released from regulatory control.
_905 ______
EU Law on Radioactive Waste
Ana Stanic E&A Law
Less than a year after the adoption of the EU Directive on nuclear safety of nuclear installations discussions are underway in the EU to adopt legislation concerning radioactive waste and spent fuel management. Specifically, the European Council (“Council”) called the European Commission (“Commission”) to continue its work towards a Community approach in this field in its Conclusions of 10 November 2009 and the European Parliament asked it to submit a new proposal for a directive.
All EU Member States generate radioactive waste. Although radioactive waste is primarily generated from the nuclear fuel programme, other activities such as use of radioactive isotopes in medicine, research and industry also give rise to such waste. Final disposal solutions are required for all existing radioactive waste, regardless of whether nuclear programmes are continued, expanded or phased out. In addition, sufficient capacity for safe interim storage facilities has to be foreseen.
At present, Community legislation dealing with spent fuel and radioactive waste deals only with the supervision and control of shipments of radioactive waste and spent fuel and nuclear safety of storage facilities for spent fuel and radioactive waste that are on the same site and are directly related to nuclear installation.
The first proposal for a Council Directive (Euratom) on the management of spent nuclear fuel and radioactive waste started was submitted by the Commission to the Council for discussion by the EC in 2003 and, resubmitted in 2004. Thereafter lengthy negotiations ensued with stakeholders in order to create a common understanding concerning the steps that need to be taken by public or private organisation to ensure the safe management of spent fuel and radioactive. At the end of 2007 the Commission initiated a 18-months study (“Study”) to analyse the collection, recording and reporting
82 Nuclear Energy f0r New Eur0pe 2010 of radioactive waste data and inventories and identify best practices and recommend measures at both national and EU levels.
By way of background this paper will discuss the conclusions of the Study in Part 1. Part 2 will examine the current proposal for the new directive and compare it to the 2004 proposal. Part 3 will detail the work currently being undertaken by the European High Level Group on Nuclear Safety and Waste Management concerning radioactive waste and spent fuel management. By way of conclusion in Part 4 the next steps in the adoption of EU legislation concering radioactive steps will be highlighted.
_906 ______
Leaching Study in Process of Solidification of Radionuclide 54Mn in Concrete
Ilija Plecas and Slavko Dimovic “VINCA” Institute of Nuclear Sciences P.Box. 522, 11001 Belgrade, Serbia and Montenegro e-mail: [email protected]
To assess the safety for disposal of radioactive waste-cement composition, the leaching of 54Mn from a waste composite into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Determination of retardation factors, KF and coefficients of distribution, kd, using a simplified mathematical model for analyzing the migration of radionuclides, has been developed. Results presented in this paper are examples of results obtained in a 30 year mortar and concrete testing project, which will influence the design of the engineered trenches system for a future central radioactive waste disposal center.
_907 ______
Teaching Activities of Centre of Experimental Geotechnics Related to Radioactive Waste Storage Based on Research Experience
Jan Smutek, Jiri Svoboda Czech Technical University in Prague, Faculty of Civil Engineering, Centre of Experimental Geotechnics Thakurova 7, 166 29 Prague 6, Czech Republic [email protected]
The Centre of Experimental Geotechnics (CEG) is a department of the Czech Technical University in Prague (CTU), part of the Faculty of Civil Engineering, concerned with both research and teaching activities related particularly to the experimental research of radioactive waste disposal.
Teaching courses deal with the problems associated with the safe isolation of radioactive waste. The CEG is involved in the PETRUS II project which concerns the development of the concept of a pan- European system of specialist training in the problems of radioactive waste disposal in deep repositories. The CTU is also a member of a number of international associations related to radioactive waste storage (ENEN, IAEA URF Net, IGD-TP, and the ITC School).
In 2007 the CEG opened the Josef Underground Educational Facility (Josef UEF) which provides not only a unique teaching facility, but also an ideal environment for research and experimentation
83 Nuclear Energy f0r New Eur0pe 2010 addressing a wide range of challenging issues in the field of underground construction. This underground workplace was created by the reconstruction of the former Josef exploration gallery which was excavated as part of the exploration of local gold-bearing deposits. Arguably the Josef UEF’s important role is to provide for in-situ research related to high-level radioactive waste disposal.
The courses provided by the CEG (in Czech and English) teach students about the basic principles of radioactive waste disposal, the characteristics of bentonite-based materials used for the construction of engineered barriers in deep underground repositories and physical modelling, the practical elements of which are taught at the Josef Underground Educational Facility. The practical teaching courses involve a combination of laboratory work and in-situ experimentation at the Josef UEF. The structure and content of the courses are based on ongoing projects in progress at the facility and the experience of CEG research and teaching personnel, which provides students with the opportunity to learn about the challenges inherent in experimental work firstly in the laboratory and then in-situ. Consequently, they gather important practical knowledge and learn firsthand about the limitations imposed by the environment in which they work.
In the last several years, CEG research projects have concentrated on radioactive waste disposal in deep underground repositories. The Czech repository concept assumes disposal in a deep geological formation (crystalline rock). It is expected that gases will eventually be created within the confines of the repository (e.g. caused by the corrosion of the waste containers which could lead to the formation of hydrogen). Such gases might absorb radionuclides from the waste and potentially, should there be leakage through the barriers, radioactivity could make its way into the geosphere. Therefore it is important to have an understanding of the gas transport properties of the host rock, a subject which has not been, to date, researched in depth. A major part of the CEG’s work is currently focused on the measurement of the gas permeability of rock masses. Ongoing projects are concerned with the study of gas transport through rock masses in the conditions expected in a deep underground repository for high-level radioactive waste.
The Josef UEF works on the principle of an open access strategy in order to build confidence in sometimes controversial issues. The general public as well as the scientific community is encouraged to view the various experiments underway at the Josef facility either personally during published visiting hours or via the internet (http://ceg.fsv.cvut.cz/, http://www.uef-josef.eu/).
_908 ______
Revision 2 of the Program of NPP Krško Decommissioning and SF & LILW Disposal
Nadja Železnik, Metka Kralj, Irena Mele, Primož Stropnik ARAO - Agencija za radioaktivne odpadke Parmova 53, Ljubljana, Slovenia [email protected]
Ivica Levanat, Vladimir Lokner, Andrea Rapić APO Ltd. Savska 49, Zagreb, Croatia [email protected]
First joint Slovenian-Croatian Program of NPP Krško Decommissioning and SF & LILW Disposal (DP) was completed in 2004 and formally adopted in 2005. As bilateral agreement on the NPP
84 Nuclear Energy f0r New Eur0pe 2010 requires periodic revisions at least each 5 years, revision 2 of DP was started in September 2008, with the purpose to incorporate relevant developments since the 1st revision, to improve the level of details and reliability of DP, and to propose updated and more accurate cost estimates and appropriate financing models.
In the first phase of the revision, new supporting studies for DP modules were prepared. Among these studies, the most demanding was the NPP Krško specific Preliminary Decommissioning Plan (PDP), complying with the IAEA-recommended format, which included development of the NPP decommissioning inventory database. For upgrade of SF management, new and more detailed descriptions with improved cost estimates were prepared. Update of LILW disposal concept was based on new developments and projects prepared for the Slovenian repository.
In the second phase of the revision, integrated DP scenarios were formulated and analyzed. They integrate NPP decommissioning together with RW and SF management/disposal into rationally inter- related sequences. Boundary conditions for this revision required: (a) that the reference scenario from the previous revision should be re-examined, with appropriate variations or new alternatives; (b) that the option of the NPP Krško life extension should also be included; and (c) that the possibility of diverging interests of the contracting parties should also be analyzed (i.e. waste division and separate management).
Finally, scenario evaluation is intended to compare the analyzed scenarios taking into account both their feasibility and estimated costs. It should provide the basis for determining future financing of DP, namely the annuities to be paid by the NPP Krško owners into the national decommissioning funds.
85 Nuclear Energy f0r New Eur0pe 2010
Nuclear Fusion and Plasma Technology
86 Nuclear Energy f0r New Eur0pe 2010
_1001 ______
Heat Transfer and Jet Interaction for Different Arrays of Impinging Jets
Martin Draksler, Boštjan Končar “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
The heat removal by multiple impinging jets is one of the most efficient cooling methods and therefore it has been selected as a cooling method for the divertor, a high-heat flux component of the future fusion reactor called DEMO. The heat removal capability is affected also by the jet arrangement in the divertor’s cooling finger.
In this study the heat transfer and flow characteristics of multiple jets impinging on the heated surface is investigated. The effect of jet-to-jet spacing in different geometrical arrangements of the jets is analyzed.
The numerical study is performed with the CFD code ANSYS-CFX, using RANS approach. Adequate turbulence models have to be used to predict the interactions between highly turbulent jets. The predictions with eddy viscosity type and Reynolds stress turbulence models will be compared and validated against the experimental data.
The resulting heat transfer will be compared for different arrays of impinging jets. The influence of jet-to-jet spacing, jet-to-wall distance and Reynolds number on the heat transfer will be considered in particular. The geometry and boundary conditions of the selected experimental cases resemble as far as possible to the DEMO divertor conditions.
87 Nuclear Energy f0r New Eur0pe 2010
_1002 ______
Calculations to Support JET Neutron Yield Calibration: Contributions to the External Neutron Monitor Responses
Luka Snoj1, Brian Syme2, Sergey Popovichev2, Igor Lengar1, Sean Conroy3 and JET EFDA Contributors* JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon, UK
1Association EURATOM-MHEST, Jožef Stefan Institute, Reactor physics division, Jamova 39, SI-1000 Ljubljana, Slovenia, [email protected]
2 EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon,OXON, OX14 3DB, UK
3 EURATOM-VR Association, Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala, Sweden
Neutron yield measurements are the basis for the determination of the absolute fusion reaction rate and the operational monitoring with respect to the neutron budget during any campaign for the Joint European Torus (JET). After the Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated 252Cf neutron source deployed inside the JET vacuum vessel. This calibration will allow direct confirmation of the JET time-resolved total neutron yield monitor calibration (which was the standard on JET originally) and provide the first direct calibration of the JET activation system (which has been the more recent standard).
A whole suite of calculations is required to support the JET neutron calibration project. Many are based on Monte Carlo modelling using the advanced Monte Carlo transport codes, such as MCNP. Here a simplified model of the JET tokamak was developed in order to study the size of various effects. Firstly a thorough study was performed to identify the main JET structures, which contribute to the external neutron monitor detector responses and to determine their neutronic importance. We have essentially verified the model postulated in the previous calibration [1] which ascribes the response to neutrons as coming mainly via the ports, with the nearest dominating and the more distant contributing according roughly to their (inverse squared) distance from the source. In addition we investigated scattering back from the torus hall wall and scattering contributions via some torus structures such as the first wall and the torus iron shell.
* See the Appendix of F. Romanelli et al., Proceedings of the 22nd IAEA Fusion Energy Conference 2008, Geneva, Switzerland [1] Numerical study of the calibration factors for the neutron counters in use at the Joint European Torus, Brian J. Laundy, Owen N. Jarvis, Fusion technology, vol.24, pages 150-160, September 1993.
88 Nuclear Energy f0r New Eur0pe 2010
_1003 ______
Development of a Large Plasma Reactor for Removal of Deposits in Fusion Reactors
Rok Zaplotnik Induktio d.o.o. Litostrojska 44D, Ljubljana, Slovenia [email protected]
Alenka Vesel, Miran Mozetič Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
A large plasma reactor that will be used as a source of oxygen or nitrogen radicals for removal of deposits in fusion reactors was constructed. The reactor is powered by a 12 kW RF generator, operating at a frequency of 27.12 MHz. The experimental system is pumped with a two stage rotary pump with a maximum pumping speed of 80 m3/h, allowing for a base pressure as low as 0.8 Pa. With the discharge chamber comprising of a 2 m long quartz tube with a 20 cm diameter, the reactor is significantly larger than usual plasma reactors used for laboratory experiments focused on the interaction of neutral radicals and fusion relevant materials. The length of the Pyrex tube will be of significant value in experiments where the reactor will be utilised as a remote source of radicals. While the detail characterisation of the plasma reactor with advanced weakly ionized plasma diagnostics (such as fiber optic catalytic probes and optical emission spectroscopy) is yet to take place, we can expect radical fluxes as high as 1021s-1.
_1004 ______
Erosion of W-C composite in Hydrogen Plasma at Temperatures Above 1000 K
Alenka Vesel, Miran Mozetič Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
Marianne Balat – Pichelin CNRS-PROMES Font Romeu-Odeillo, France
Rok Zaplotnik Induktio d.o.o. Litostrojska 44D, Ljubljana, Slovenia
Plasma facing components (PFC) of fusion devices are due to intensive particle and heat loads subjected to high material erosion, migration and re-deposition. Especially graphite tends to be eroded by hydrogen ions during hot plasma pulses. The eroded material deposits on different walls, especially at remote parts of fusion reactors, and they represents an unpredicted source of hydrogen. Furthermore, these deposited layers may be subjected to further re-erosion. The eroded wall material can be transported into the core plasma where they cause dilution of the plasma fuel and cooling of the plasma itself and this reduces the effectiveness of the fusion process. Therefore, the formation of re- 89 Nuclear Energy f0r New Eur0pe 2010 deposited layers of hydrocarbons which can be contaminated also with some metals (W) is of major concern in the development of the next-step fusion devices.
Here we present a study on interaction of neutral hydrogen atoms with W-C composite. 1 m thick W-C composite was deposited on stainless steel substrate and treated in microwave hydrogen plasma. Treatment time was kept constant at 5 min. During plasma treatment additional sample heating was performed by the use of concentrated solar radiation. Temperature of the samples was varied between 1000 K and 1280 K. The temperature was measured by pyrometer. This allowed us to follow the surface chemical modification occurring during plasma treatment. A change in the chemical status of the surface caused deviation in the sample temperature due to changes in the sample’s emissivity. After the treatment the samples were analyzed by classical techniques for surface characterization like x-ray diffraction (XRD), x-ray photoelectron spectroscopy (XPS), Auger electron spectroscopy (AES) and scanning electron microscopy (SEM). The results have shown the reduction of carbon and segregation of iron from stainless steel substrate to the surface and formation of W-Fe alloy, probably
Fe7W6.
_1005 ______
Travelling Exhibition – Fusion Expo
Melita Lenošek, Saša Novak Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
The Fusion Expo is an itinerant exhibition presenting various aspects of fusion research such as: fusion as a natural phenomenon and energy source, fusion as a European research project, history of fusion research, European research facilities, ITER, plans toward development of a fusion power plant and technological, environmental and sociological aspects of this energy source. Main target group of this travelling exhibition is the general public, specifically designed to address young students. It is also suitable for addressing other audiences, such as decision makers or journalists.
Since beginning of Fusion Expo support actions under EFDA in 2008, the Fusion Expo has been the responsibility of the Slovenian Fusion Association [1]. Since then we are trying to bring it to visitors on most interactive way.
The paper will present past experiences, some findings, improvements and challenges of Fusion Expo support actions under EFDA (European Fusion Development Agreement) [2].
REFERENCES [1] http://www.sfa-fuzija.si [2] http://www.efda.org/index.htm
90 Nuclear Energy f0r New Eur0pe 2010
_1006 ______
Microstructure and Mechanical Properties of SiC Fibers for Potentional Use in a Future Fusion Reactor
Tea Toplišek, Goran Dražić, Spomenka Kobe “Jožef Stefan” Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
Vilibald Bukošek Faculty of Natural Science and Engineering Snežniška 5, SI-1000 Ljubljana, Slovenia [email protected]
A monolithic silicone carbide (SiC) has been recognized as one of the most promising structural materials for many thermo-mechanical applications because of its excellent high-temperature strength and modulus, good oxidation resistance, high hardness, low specific weight and low density. The problem with monolithic SiC is its low thermal shock resistance, which leads to cracking and catastrophic failure of the material. This can be improved by introducing a reinforcement phase, continuous SiC fibers, to produce a SiCf/SiC composite material. This kind of composite is being considered for a future fusion reactor because of its low induced radioactivity after neutron irradiation, non-catastrophic failure mode, specific thermal conductivity and low porosity.
In this work the surface, morphology, topography and chemical composition of different SiC fibers was observed and characterized using scanning (SEM) and transmission (TEM) electron microscope and atomic force microscope (AFM).
The mechanical properties of SiC fibers were measured on dynamometer Instron 5567. Because SiC fibers exhibit a wide distribution of diameters each fiber was optically checked to get a reliable tensile strength data. The fibers were glued onto paper in the shape of letter “U” to make sure; that the presence of the creep inside the clamps was eliminated. With special program we can follow internal changes in material structure during testing. A clamping length was 1 cm and a crosshead speed of 1 mm/min.
After tensile measurements all fracture surfaces of the fibers were investigated with scanning electron microscope. In general, fracture in fibers initiates at some flaw(s), internal or on the surface. Very frequently, a near-surface flaw such as a microvoid or an inclusion is responsible for the initiation of fiber fracture. Surface flaws are common in SiC based fibers because of the processing technique. Airborne particles as well as other elements tend to attach to the surface of the fiber during process and handling. One important feature of ceramic fiber is the surface texture. Their surface roughness scales with grain size. The rough surface of such brittle fibers makes them break at very low strains and it makes very difficult to handle them in practice. The grain boundaries on the surface can act as notches and weaken the fiber.
Another mechanical property measured for SiC fibers was the nanohardness using Vickers indenter on the Fischerscope instrument.
91 Nuclear Energy f0r New Eur0pe 2010
_1007 ______
Calculations to Support JET Neutron Yield Calibration: Neutron Scattering in Source Holder
Luka Snoj1, Brian Syme2, Sergey Popovichev2 and JET EFDA Contributors*
JET-EFDA, Culham Science Centre, OX14 3DB, Abingdon, UK
1Association EURATOM-MHEST, Jožef Stefan Institute, Reactor physics division, Jamova 39, SI-1000 Ljubljana, Slovenia, [email protected]
2 EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon,OXON, OX14 3DB, UK
In 2010 the Joint European Torus (JET) plasma facing wall will be changed from a carbon wall to an ITER-Like Wall (Beryllium/Tungsten/Carbon). After that transition, the confirmation of the JET neutron yield calibration will be ensured by direct measurements using a calibrated 252Cf neutron source deployed inside the JET vacuum-vessel. The neutron source will be deployed on the JET Mascot robot. In order to safely manipulate the neutron source with the robot, the source will be placed in a specially designed source holder (source baton), which will connect to the mascot robot via a longer tube, called the mascot baton. The two batons approach is needed in order to ensure adequate separation of neutron source from the robot body, in order to reduce neutron scattering from the mascot robot, to reduce activation of the robot and to limit the dose on the robot cameras.
The source baton is subject to several design constraints; it should retain its structural integrity in the event of accident, the connection of the source and mascot baton should be fail safe and the baton should not significantly disturb the spatial flux distribution and energy spectrum of the neutron source. The latter point is especially important for accurate calibration. Hence a set of calculations is needed to support the baton design.
The paper examines the effect of various baton design features on neutron flux distribution and neutron spectra. Using the final baton design, we present a thorough study of the effect of the baton on angular neutron flux distribution around the source holder, on neutron spectrum and on JET activation detector response. All calculations were performed by using the Monte Carlo code MCNP [1].
* See the Appendix of F. Romanelli et al., Proceedings of the 22nd IAEA Fusion Energy Conference 2008, Geneva, Switzerland [1] X-5 Monte Carlo Team, MCNP - A general Monte Carlo N-particle Transport code, Version 5, LA-UR-03-1987, April 24, 2003 (revised June 30 2004).
92 Nuclear Energy f0r New Eur0pe 2010
_1008 ______
A Fully-kinetic PIC Simulation of an Emissive Probe in Tokamak Relevant Plasma
J. Kovačič1, T. Gyergyek1,2, M. Čerček2,3
1 University of Ljubljana, Faculty of Electrical Engineering Tržaška cesta 25, SI-1000 Ljubljana, Slovenia [email protected]
2 “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia
3 University of Maribor, Faculty of Civil Engineering Smetanova 17, SI-2000 Maribor, Slovenia
In our work we made an attempt at simulating an emissive surface e.g. emissive probe via fully-kinetic PIC simulation. This is an important issue, which deals with the feasibility of using an emissive probe (also laser emissive probe) in tokamak plasma [1, 2]. The particle-in-cell computer simulation model is an electrostatic 1d3v system, bounded on both sides by infinite floating electrodes, with a volumetric source in the middle of the simulation domain. With the use of artificial heating to maintain a required Maxwellian velocity distribution of the particles in the source region each electrode could have a different value of the electron emission current. Maxwellization of the source particles velocity distribution function is also important to simulate a realistic plasma source. We also incorporated many realistic mechanisms of the edge layer of the plasma such as self-consistent collisions of charged and neutral particles, secondary emission, magnetic field, production of neutrals at the walls. All of this was done using massively parallel particle-in-cell code BIT1 [3], which scales up excellent on large high performance computers. Since the system had to be rather long to eliminate the effects of the neutrals in the source region, the simulation proved to be very demanding regarding CPU time. In our simulation runs we simultaneously used up to 2000 processors for various sets of parameters, with each simulation run taking up to a few days. The results were qualitatively compared with some emissive probe measurements. [1] R. Schrittwieser et al., “Measurments with an emissive probe in the CASTOR tokamak”, Plasma Phys. Control. Fusion, 44, (2002), 567-578
[2] R. Schrittwieser et al., “A radially movable laser-heated emissive probe”, J. Plasma Fusion Res., 8, (2009), 632-635
[3] D. Tskhakaya et al., “PIC/MC code BIT1 for plasma simulations on HPC”, Proceedings of the 18th Euromicro Conference on the Parallel, Distributed and Network-based Processing, Pisa, Italy, (2010), 476-481, IEEE, Computer Society, editors M. Danelutto, J. Bourgeois, T. Cross
93 Nuclear Energy f0r New Eur0pe 2010
_1009 ______
Vibrationally Excited Hydrogen Molecules from Desorptive Recombination at Surfaces
Vida Žigman University of Nova Gorica Vipavska cesta 13, SI-5000 Nova Gorica, Slovenia [email protected]
A kinetic model for describing hydrogen - wall interaction, simultaneously predicting volume population and surface adsorbed population of particles is put forward. Four particle species are taken into account: besides volume hydrogen atoms and molecules in ground state, vibrationally excited hydrogen molecules and surface adsorbed hydrogen atoms are both treated as distinct particles. Particle-surface interactions with both hot (~1800 K) and cold (300 K) walls of the container are considered. The main are taken to be: vibrational excitation and deexcitation, and thermodissociation on the hot walls (filament); and vibrational deexcitation and recombination on the cold walls. It is the latter interaction that we are specially concerned with, considering desorptive recombination through the Eley-Rideal process.
Depending on the surface bond mechanism of the adsorbed atom, physisorption or chemisorption, the interaction with the impinging bulk gas atom results in a molecule in an excited vibrational state or in ground state, respectively. The adsorptive behaviour is taken accordingly to Henry's law, assuming that atoms adsorb to the surface independently of one another. The model is self-consistent, the system of four coupled rate equations is closed. Data on surface processes (transition probabilities, sticking coefficients, etc…) are needed and the inverse processes' probabilities are taken according to the detailed balance principle.
The improvement along the assumption of having only a finite number of surface adsorption sites, which gives a coverage in form of a monolayer is in progress.
_1010 ______
ITER, an Essential Step in Fusion Development
Jesus Izquierdo European Joint Undertaking for ITER and the Development of Fusion Energy, C. Josep P. 2, B3, 08019 Barcelona, Spain [email protected]
ITER in Latin means the way. ITER is the world’s largest scientific partnership that brings together seven parties that represent half of the world’s population- the EU, Russia, Japan, China, India, South Korea and the United States. The aim is to demonstrate the feasibility of fusion as a sustainable source of energy.
Since 2007, year in which ITER Organization entered into force, a number of milestones for the construction of this ‘first of a kind’ nuclear installation has been achieved. Fusion for Energy (F4E), the European Domestic Agency for ITER, has been playing a key role in this achievements: more than one hundred procurement procedures to manufacture ITER components have been launched: the manufacturing of the Vacuum Vessel, the supply of copper and Niobium-tin (Nb3Sn ) strand, the Architect Engineering for the buildings or the fabrication of its seismic isolators.
94 Nuclear Energy f0r New Eur0pe 2010
In this talk, a summary of the progress of the project will be presented with emphasis in the main challenges that the project is facing from the scientific and technological point of view.
_1011 ______
Particle-In-Cell (PIC) Simulations and Grid-Free Treecode Method
Janez Krek LECAD Laboratory, Faculty of Mechanical Engineering Aškerčeva 6, SI-1000 Ljubljana, Slovenia [email protected]
Nikola Jelić Association EURATOM-ÖAW, Institute for Theoretical Physics, University of Innsbruck, Technikerstraße 25, A-6020 Innsbruck, Austria [email protected]
Jože Duhovnik LECAD Laboratory, Faculty of Mechanical Engineering Aškerčeva 6, SI-1000 Ljubljana, Slovenia [email protected]
The grid-free treecode (TC) method is well known method for calculating interactions between bodies and/or particles in various systems [1]. In terms of accuracy and speed of method, treecode method can be positioned between direct integration, with typical process of O(N2) and PIC method [2], with typical process of O(N log N).
In this work we present basic principle of treecode method, its implementation in main PIS simulation loop and its advantages and weaknesses. Algorithm of treecode method shows capabilities to efficiently run code in multi-processors (or multi-core) computers and we tried to implement this capability in our treecode program. We created small but efficient program in C language to fully present, show and test treecode method on field of PIC simulation.
In results, we compared running times for program compiled for single-processor with program compiled for multi-processor and also compared the results for treecode for simple 1-D case of virtual cathode with results given in [3] with program in MatLab. Refernces:
[1] J. Barnes and P. Hut. A hierarchical o(nlogn) force-calculation algorithm. , 324:446–449, dec 1986.
[2] C. K. Birdsal and A. B. Langdon. Plasma Physics via Computer Simulation. McGraw-Hill, 1985.
[3] Andrew J. Christlieb, R. Krasny, and John P. Verboncoeur. Efficient particle simulation of a virtual cathode using a grod-free treecode poisson solver. IEEE Transactions on plasma science, 32(2):384–389, April 2004.
95 Nuclear Energy f0r New Eur0pe 2010
_1012 ______
Focused 3He Ion Beam: Highly Selective and Laterally Resolving Method for Deuterium Detection in Plasma-Facing Components
Primož Pelicon, Primož Vavpetič, Zdravko Rupnik, Iztok Čadež, Sabina Markelj, Nataša Grlj, Mirko Ribič, Zvonimir Grabnar Jožef Stefan Institute, Jamova 39, Association EURATOM-MHEST, P.O.B. 3000, SI-1001 Ljubljana, Slovenia [email protected]
Nuclear Reaction Analysis with 3He ion beam is a powerful analytical technique for analysis of light elements in thin films and surface layers. Main motivation for the 3He focused beam application at JSI tandetron accelerator is lateral mapping of deuterium using the nuclear reaction D(3He,p)4He. Due to the complexity in handling the tritium-contaminated objects from JET, the evaluation of the fuel retention processes in wall materials rely in big part on deuterium quantification after deuterium plasma exposure experiments in other European tokamaks. As the retained fuel in plasma-facing wall components exhibits strongly inhomogeneous lateral distribution, the method resolves the features at the micrometer scale and in this way provides unique information for fuel retention studies.
The He beam at the tandetron ion accelerator of JSI is formed inside the duoplasmatron ion source and extracted in the negative charge state from the lithium charge exchange channel. The injection of the beam in the JSI microprobe results in the typical ion currents of 300 pA and the beam dimension of 4 x 4 µm2. We will present the 4.5 MeV 3He2+ focused beam formation at JSI. The NRA detection setup consists of a charge-particle implanted silicon detector used in parallel with the permanently installed X-ray detector, RBS detector and a beam chopper for ion dose monitoring. The method to quantify deuterium absolutely is described. Plasma-deposited amorphous deuterated carbon thin films (a-C:D) with known D content were used in addition as a reference [1].
The method was used to study deuterium fuel retention in Carbon Fiber Composite (CFC) materials exposed to deuterium plasma in Tore Supra [2] and TEXTOR controlled fusion devices. Lateral deuterium mappings of plasma-facing surfaces, inside the gaps of castellation as well as along the cleavages perpendicular to the exposed surfaces have provided important information for the application of CFCs in a new generation of thermonuclear reactors.
References
[1] P. Pelicon, P. Vavpetič, N. Grlj, I. Čadež, S. Markelj, S. Brezinsek, A. Kreter, T. Dittmar, E. Tsitrone, M. Rubel and T. Schwarz-Selinger, “Deuterium mapping with micro-NRA using high energy focused 3He beam for fuel retention studies in the walls of fusion reactors”, Conference ICNMTA 2010, Leipzig. Accepted abstract. To be published in NIM B.
[2] T. Dittmar, E. Tsitrone, B. Pégourié, I. Čadež, P. Pelicon E. Gauthier, P. Languille, J. Likonen, A. Litnovsky, S. Markelj, C. Martin, M. Mayer, JY Pascal, C. Pardanaud, V. Philipps, J. Roth , P. Roubin and P. Vavpetič, “Deuterium Inventory in Tore Supra: fuel retention in the gaps”, Presented at the Conference PSI 2010, San Diego. Submitted to Journal of Nuclear Materials.
96 Nuclear Energy f0r New Eur0pe 2010
_1013 ______
Hydrogen Permeability of Beryllium Films
Bojan Zajec, Vincenc Nemanič “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected] C. Porosnicu, C.P. Lungu National Institute for Lasers, Plasma and Radiation Physics, NILPRP Magurele-Bucharest 077125, Romania
Beryllium films will be applied in fusion reactors as one of the main components (Be, W, CFC) of the surface layer first wall in the near future [1]. The morphology, adhesion and thermal stability of thin films deposited by the thermionic vacuum arc (TVA) method have been confirmed to be compatible with the required extremely high demands [1, 2]. The role of Be in tritium retention and release may be well predicted assuming that low solubility of hydrogen in Be suppresses migration of the fuel into the tiles. Actual Be films used in the future experiments may contain voids or channels causing unwanted porosity which may change the tritium interaction kinetics with the coated tiles.
We present results of precise hydrogen permeation measurements through 8 micrometer thick Be films deposited by TVA at NILPRP on Eurofer steel. Substrates were 0.5 mm thick 40 mm diameter discs having the geometric area exposed to hydrogen of 8.4 cm2. The permeation reduction factor (PRF) of Be coated membranes at 400 °C and upstream hydrogen pressure of 1 bar varied roughly from 15 to 2000 compared to the uncoated Eurofer membrane. Such a large scatter of results over 7 samples deposited at identical condition was not expected. Fast transients of the observed permeation flux which followed the sudden upstream pressure change supports the picture that Be films are porous on the micro-scale or nano-scale. The size or distribution of the leak channels that would support the observed behaviour could be predicted by theoretical models. Inter-granular micro-channels were indeed revealed on low-angle cross sections of the Be films by SEM. The PRF of a particular membrane could be additionally modified when the Be films were intentionally oxidized. The subsequent XPS analysis shows that a thin BeO layer is grown on the Be surface, but it should always increase the PRF. The process responsible for permeation flux change is presumably the oxidation of the Eurofer membrane inside the Be micro-channels.
References:
[1] G F Matthews, P Edwards, T Hirai, M Kear, A Lioure, P Lomas, A Loving, C P Lungu, H Maier, P Mertens, D Neilson, R Neu, J Pamela, V Philipps, G Piazza, V Riccardo, M Rubel, C Ruset, E Villedieu and M Way on behalf of the ITER-like Wall Project Team 1−11, Overview of the ITER-like wall project, Phys. Scr. T128, 2007, 137.
[2] C. P. Lungu, I. Mustata, V. Zaroschi, A. M. Lungu, A. Anghel, P. Chiru, M. Rubel, P. Coad G. F. Matthews and JET-EFDA contributors, Beryllium Coatings on Metals: Development of Process and Characterizations of Layers, Phys. Scr. T128, 2007, 157
97 Nuclear Energy f0r New Eur0pe 2010
_1014 ______
Mechanical Properties and Microstructural Characterizations of Potassium Doped Tungsten
Hua Sheng, Inge Uytdenhouwen, Vincent Massaut SCK•CEN, the Belgian Nuclear Research Centre, 2400 Mol, Belgium [email protected], [email protected], [email protected]
Guido Van Oost Department of Applied Physics, Ghent University J.Plateaustraat 22, B-9000 Ghent, Belgium [email protected]
Jozef Vleugels Department of Metallurgy and Materials Engineering, K.U.Leuven Kasteelpark Arenberg 44, B-3001 Leuven, Belgium [email protected]
Tungsten is a very promising candidate material for plasma facing components in fusion reactor due to its high melting temperature, high thermal conductivity, low tritium inventory and low erosion rate under plasma loading. The main drawbacks are its poor mechanical properties. None of the W & W alloys developed so-far however has been fully optimized for structure or armour application in fusion reactors. Nor have reference tungsten grades been fully characterized.
The tensile properties of tungsten in the vicinity of room temperature, the ductile-brittle transition must be considered because of its effect on yield strength and other mechanical behaviours. The potassium doped tungsten grade WVWM (as received and annealed at 1800°C) produced by Plansee AG could be a potential plasma facing material for future nuclear fusion facilities and reactors such as ITER and especially DEMO.
For a better understanding of both recrystallization and ductile to brittle transition temperature, tensile tests are performed on potassium doped tungsten, WVWM, up to 2000°C at different loading rates (0.2 and 42 mm/min). The mechanical properties are highly dependent on the microstructure. The etched sample surface and fracture surface after tensile testing are microstructurally assessed by Optical Microscopy (OM), Scanning Electron Microscopy (SEM) and Transmission Electron Microscopy (TEM). All the obtained data will be summarized into a database that can be used for future modelling input and will be compared to the same grade material after neutron irradiation.
98 Nuclear Energy f0r New Eur0pe 2010
_1015 ______
A Monte Carlo model of the D-D neutron source at FNG
Alberto Milocco1), Andrej Trkov1), Mario Pillon2), Roberto Bedogni3) 1)Jožef Stefan Institute Jamova cesta 39, SI-1000 Ljubljana, Slovenia 2)ENEA FPN-Fusione, Via Enrico Fermi, 45-00044 Frascati, Rome, Italy 3)Istituto Nazionale di Fisica Nucleare, Via Enrico Fermi, 40-00044 Frascati, Rome, Italy Corresponding author: [email protected]
The Deuteron – Deuteron (D – D) fusion reactions are foreseen for the next stage fusion reactors. The Frascati Neutron Generator (FNG), located near Rome, can produce neutrons of about 3 MeV by accelerating Deuterons onto solid targets containing Deuterium A computatational model is developed for the simulation of the FNG neutron source. It requires the specification of the beam and target characteristics (e. g. Deuterons energy, Deuterium atomic fraction). The model is implemented in the subroutines of the MCNPX and MCNP5 codes, which need so far to be recompiled. The Deuterons are transported inside the solid target by a Monte Carlo method. The neutrons are generated with the angle – energy distribution as defined by the laws and nuclear data for the Deuteron – Deuteron reaction in the ENDF/B-VII library. The sensitivity studies on the input parameters of the D – D model are presented. The D – D source model is validated with an experiment, which has been performed by the INFN team at the FNG with a set of Bonner spheres and the D – D source. The experimental response functions are simulated. An unfolding technique is used to reconstruct the neutron source spectra. These are compared with the spectra calculated by the D – D model. The anisotropy properties of the source are also addressed. In conclusion, the modelling of FNG experiments with the D – D neutron source is feasible concerning the main features and associated uncertainties of the source term.
_1016 ______
A Fluid Model and PIC Simulation of Two-Electron-Temperature Plasma in an Oblique Magnetic Field
J. Kovačič1, T. Gyergyek1,2, M. Čerček2,3
1 University of Ljubljana, Faculty of Electrical Engineering Tržaška cesta 25, SI-1000 Ljubljana, Slovenia [email protected]
2 “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia
3 University of Maribor, Faculty of Civil Engineering Smetanova 17, SI-2000 Maribor, Slovenia
The objective of our work was to make a model of collisionless plasma with an additional population of hot electrons, which would then be studied in the oblique magnetic field. Similar problem was studied for plasmas with negative ions by [1]. Our model is a modified Tonks-Langmuir model with ions that are born with negligible velocity and two electron populations with different temperatures. The parameters that are varied are the ratios of electron densities and electron temperatures as well as the angle and density of the magnetic field. A formation of double layers was studied along with the occurrence of the magnetic presheath. We obtained the profiles of various plasma quantities via 99 Nuclear Energy f0r New Eur0pe 2010 numerical integration of the system of differential equations. We occasionally had to use an implicit scheme for Newton method due to the stiffness of the problem. A model is flexible in a way that you can normalize quantities either to cold or hot electrons and observe the system in two different “scales”. We had also set up an equivalent plasma model in the PIC simulation code BIT1 [2], which enables the maintaining of the Maxwellian velocity distribution, one of crucial details of the model. The length of the analytical and the simulation system is connected through ionization lengths and the results obtained with the two methods are therefore not straightforward to compare. [1] T. E. Sheridan, »Double layers in a modestly collisional electronegative discharge«, J. Phys. D: Appl. Phys., 32, (1999), 1761-1767.
[2] D. Tskhakaya, S. Kuhn, »Kinetic (PIC) simulations of the magnetized plasma-wall transition«, Plasma Phys. Control. Fusion, 47, (2005), A327-A337.
_1017 ______
The Investigation of Fuel Energy Gain for Tritium-Poor Fuels in Fast Ignition Fusion
M. Moosavi, A. Ghasemizad P.o.Box.41335-1914, Rasht, Iran [email protected] , [email protected]
The fast ignition scheme is one of the most fascinating and feasible ignition schemes for the inertial fusion energy. Fast ignition targets are ignited on the outside surface so they have no need for a low density, high temperature center required by central hot spot ignition. The fast ignition (FI) concept requires the generation of a compact, dense, pure fuel mass accessible to an external ignition source. In the fast-ignitor concept the capsule is imploded by a conventional laser to produce a high-density core; there, the core is ignited using a short-pulse laser with high intensity.
This idea stirred immediate and intense interest around the world for two reasons. First, the approach removes the requirement to drive a series of shocks through the compressed fuel to form the spark at the center, allowing more fuel to be compressed for the same laser drive energy. Since the fusion- energy gain depends only on the amount of fuel present, once the spark has been formed, the energy gain rises by a factor of 300 rather than the factor of 15 obtained with indirect Hohlraum drive, as expected from the NIF.
Second, if more fuel can be compressed during the implosion, then the in-flight aspect ratio (the ratio of the thickness of the remaining shell and fuel during the acceleration phase of the implosion to the initial shell radius) can be much smaller than for the conventional scheme. This, in turn, substantially reduces the implosion symmetry needed to assemble the fuel to high density, because the growth rate of hydrodynamic instabilities that deleteriously mix the fuel and shell materials is known to significantly increase as the in-flight aspect ratio rises.
This is in contrast to the isobaric hot-spot scenario, where it is essential that the hot-spot area and the surrounding main fuel remain in pressure equilibrium during compression. In Fast ignition concept, isochoric configuration ignites with a laser beam.
A particularly interesting aspect of this study is the use of a target with very small total fraction tritium content Ft and net tritium breeding. Many advantages have been claimed for tritium-poor fuels: lower
100 Nuclear Energy f0r New Eur0pe 2010 tritium inventories, lower tritium-breeding requirements, lower neutron wall loadings and the possibility of direct energy conversion cycles.
In this paper, we calculated fuel energy and fuel energy gain for a capsule nearly pure deuterium, ignited by a DT seed, which would reduce the tritium inventory to a few percent. The results of the investigation are presented and discussed. Energy gain curves for different tritium concentrations are found, and limiting gain curves are derived. Finally, the potential of tritium-poor fast ignitors is then compared with that of equimolar DT fast ignitors. Refrences:
[1] S. Atzeni, M.L. Ciampi, Nul. Fusion, Vol. 37, No, 12 (1997)
[2] ''An Introduction to Inertial Confinement Fusion'', S. Pfalzner.(2006)
[3] A.I. Mahdy H. Takabe, K. Mima, Nul. Fusion, Vol. 39, No. 4(1999)
[4] C. Deutsch, “Fast ignition schemes for inertial confinement fusion”, Eur. Phys. J. Appl. Phys. 24, 95–113 (2003)
_1018 ______
The Shape of the Potential Profile Near the Boundary in the Tonks-Langmuir Model to the Case of Finite Ion–Source Temperature
Leon Kos, Jože Duhovnik University of Ljubljana, Faculty of Mechanical Engineering Aškerčeva 6, SI-1000 Ljubljana, Slovenia [email protected], [email protected]
Nikola Jelić Association EURATOM-ÖAW, University of Innsbruck, Department of Theoretical Physics, A-6020 Innsbruck, Austria [email protected], University of Ljubljana, Faculty of Mechanical Engineering SI-1000 Ljubljana, Slovenia [email protected]
The problem of the potential profile shape near the plasma boundary is an important one in the plasma theory and its application. Complete numerical plasma and sheath equation was first obtained by Self [1] many years ago, for the case of zero initial ion-source temperature, while the case of finite ion- source temperatures was attacked by Robertson [2] in 2009 only on equidistant computational grid with rather low resolution. The solution with a considerable increased number of computational cells with at a flexible grid enabling extremely high resolution as approaching the wall was obtained also in the last year by Kos and Jelić et al. [3, 4]. Therefore the problem seems to be completely overcame. However, in spite of this fact, approximate analytic solutions for plasma region and sheath region still remains of high interest for plasma physics. This is not only for theoretical but also for pragmatic point of view because this approach helps us to define a reasonable plasma-sheath transition via constructing analytic solution which smoothly patches plasma and sheath branches of curves.
It has been shown by Riemann [5] that in kinetic approach this problem can be solved successfully only for cold ion sources but that in the case of finite ion-sources the problem is extremely stiff
101 Nuclear Energy f0r New Eur0pe 2010 because of mathematical structure of the basic integral-differential equation which does not permit any reasonable approximate solution. In order to, anyway, not gave up we proceed with a particular difficulty as we describe as follows. Namely the first step towards finding approximate sheath solution is necessary is to find the value of coefficient alpha with corresponding coefficient in the form of the α sheath edge singularity (Φs-Φ) → C(x-xs) describing the limiting potential variation in front of the sheath edge (xs, Φs). Since for the cold ion-source case analytic solution is not difficult to obtain we known safely that alpha in this case is exactly ½. For finite ion-source Riemann expects α=2/3, but this was never proven. Via using our numerical solution applied to a large number of ion-source temperatures we show that Riemann's expectation [6] is correct except in a very narrow region. Our result will enable one to proceed with “impossible mission” of finding proper plasma, sheath and finally intermediate region analytic expression.
REFERENCES
S. A. Self, “Exact Solution of the Collisionless Plasma-Sheath Equation”, Phys. Fluids, 6, 1963, pp. 1762.
S. Robertson, “Sheath and presheath in plasma with warm ions”, Phys. Plasmas, 16, 2009, pp. 103503 .
L. Kos, N. Jelić, S. Kuhn, and J. Duhovnik, “Extension of the Bissel-Johnson plasma-sheath model for application to fusion-relevant and general plasmas”, Phys. Plasmas, (9) 16, pp. 093503.
N. Jelić , L. Kos and D.D. Tskhakaya sr. “The ionization length in plasmas with finite temperature ion sources”, Phys. Plasmas, 11, 2009, pp. 4158-4166
K.-U. Riemann, “The Bohm criterion and sheath formation”, J. Phys. D: Appl. Phys., 24, 1991, pp. 493.
K.-U. Riemann, “Different models of the plasma-sheath transition”, 62nd Annual Gaseous Electronics Conference, APS Meeting Abstracts, American Physical Society, 2009, pp. 1001.
_1019 ______
SITE-P: A Novel Route for Preparation of SiCf/SiC Composites
Aljaž Ivekovič, Saša Novak, Goran Dražić Department for Nanostructured Material, Jožef Stefan Institute, Slovenia Slovenian Fusion Association (SFA) EURATOM-MHEST
Fabrication of the continuous SiC-fibre reinforced SiC-ceramics for fusion applications represents a major challenge for material science, since the material to be used in the first wall of the future reactor has to meet several highly demanding requirements. Beside the appropriate mechanical properties and stability at the operating temperature, i.e. up to 1100 °C, limited closed porosity and low neutron activation, one of the main issues remained sufficient thermal conductivity. Namely, the state-of-art techniques fail in meeting at least one of the required material’s characteristics.
Recently we introduced “SITE” process as an optional way that comprises ceramic slip-infiltration in SiC-fibre perform, followed by moderate-temperature densification of SiC-based matrix using transient eutectoid. It was found, however, that the thermal conductivity of the resulting material allows the use in flow channel inserts, but it was too low for the structural applications. Therefore, as
102 Nuclear Energy f0r New Eur0pe 2010 another alternative we have introduced a novel route, where a pre-ceramic polymer is introduced into the green ceramic body after the infiltration of SiC-fibre preform with SiC powder suspension. In the next step, the infiltrated part is thermally treated to pyrolyse the polymer precursor and crystallise the newly formed SiC. The resulting material is relatively pure SiC.
In the present paper, we will present the processing route to form SiCf/SiC composite by the so called SITE-P technique and some of the materials properties. It will be presented, that the thermal conductivity of the final material even exceeds the required properties, while the decreasing of the porosity still needs further efforts.
_1020 ______
Compatibility of W-core β-SiC Fibers with SiC based Composite Material for Fusion Application
Goran Dražić, Tea Toplišek, Aljaž Ivekovič, Saša Novak “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected], [email protected]
SiC based composite is practically the only non magnetic material that could be used in structural applications in fusion environment (low activation in neutron flux, operating temperature higher than ° 1000 C and radiation-defects resistant). It should be in a form of 3D SiCf/SiC composites with high thermal conductivity and mechanical strength normal to the thickness. The SiC should be in cubic (beta) form with very small amount of porosity and gas-tight.
According to the present state of the art, materials prepared by CVI do not meet the required properties and no clear adequate improvement route in material science or technology to meet these requirements has been proposed. Among few important properties which are not achieved is the thermal conductivity of the material.
One possible solution for this problem is introduction of tungsten filaments in the material with its intrinsic room-temperature thermal conductivity of 170 W/mK. By incorporation of W into SiC-based matrix in a proper amount and geometry, the target through thickness thermal conductivity of >30 W/mK should be achieved.
The aim of the work was to study the possibility of incorporation of SiC-coated tungsten fibres, compatible with the fabrication process for production of dense SiC-based matrix and compatible with the conditions at which the material will be used. As model materials different grades of SiC-coated W-wires/fibres from TISICS Ltd, UK were used.
Among the main concerns is the potential instability of W in contact with SiC at high temperatures. Based on the literature data and preliminary investigations performed we found that reaction products may have a detrimental effect on mechanical properties.
In the work the detail investigation of the W-SiC stability at the material processing (using SITE process) and at operating conditions, as well as the analysis of mechanical properties affected by the formation of reaction layer were described and discussed.
103 Nuclear Energy f0r New Eur0pe 2010
_1021 ______
Production of Vibrationally Excited Hydrogen Molecules from Tungsten
Sabina Markelj, Iztok Čadež, Zdravko Rupnik “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
Production of hydrogen molecules by ion and atom surface recombination is an important process for hydrogen recycling on the walls of fusion reactor. Processes involving hydrogen atoms and molecules are very important for plasma behaviour close to the walls of reactor especially in the region of tokamak divertor. Plasma is there intentionally directed to the walls, therefore power loads to the divertor targets are many times bigger than to the rest of the walls. In order to minimise the power load to the divertor targets, the power has to be spread over larger volume. This is achieved by producing a neutral gas cushion where hydrogen atoms and molecules have an important role.
In order to study production of hydrogen molecules by atom recombination on surfaces, we have made a new set-up where vibrational population of hydrogen molecules emitted from the studied surface can be determined. A sample made of the studied material is exposed to a well defined beam of hydrogen atoms created by hydrogen atom beam source HABS [1]. Sample is mounted on a holder whose temperature can be varied from 10oC to 400oC. Produced molecules are analysed by special spectrometer [2], [3], where vibrational population in the emitted gas 6.5 cm from the surface is determined. We have studied production of vibrationally excited hydrogen molecules emitted from polycrystalline tungsten surface as a function of sample temperature. Results will be presented and compared to the previous measurements on tungsten by different set-up [4].
[1] http://www.mbe-components.com/products/gas/habs.html; K.G. Tschersich, J.P. Fleischhauer and H. Schuler, J.Appl.Phys. 104 (2008) 034908
[2] S. Markelj, I. Čadež, Z. Rupnik, Int. J. Mass Spectrom 275 (2008) 64-74
[3] I. Čadež, S. Markelj, Z. Rupnik, P. Pelicon, Journal of Physics: Conference Series 133 (2008) 012029
[4] I. Čadež, S. Markelj, P. Pelicon, Z. Rupnik, J. Nucl. Material 390-391 (2009) 520
104 Nuclear Energy f0r New Eur0pe 2010
Radiation and Environment Protection
105 Nuclear Energy f0r New Eur0pe 2010
_1101 ______
Discharged Radionuclides from the Slovenian Nuclear & Radiation Facilities and their Actual Detection in the Environment
Milko J. Križman, Michel Cindro, Barbara Vokal-Nemec Slovenian Nuclear Safety Administration Železna 16, SI-1000 Ljubljana, Slovenia [email protected], [email protected], [email protected]
Operation of nuclear and radiation facilities is inevitably bound to discharge radioactive materials to the environment. The control of discharged radionuclides and monitoring of their concentrations in the environment are important issues of the operation licence. Measurements of additional radioactivity in the environment are the key for the assessment of public exposure. Some radionuclides that are released in higher amounts can easily be detected in the environmental media, even though they are dispersed and diluted. Actually, it is not possible to identify radionuclides that are discharged to the environment, in extremely low activities. Their detection depends mostly on the quantity of environmental samples, the sensitivity of applied methods and the quality of the analytical equipment.
According to the legal classification, there are three nuclear installations in Slovenia, the Krško Nuclear Power Plant, the research reactor TRIGA and the interim low and intermediate radioactive waste storage, all of them in operation. The Krško NPP is currently releasing the highest activities of radionuclides, mostly 3H, 14C, activation and fission products. At this moment, the only radiation facility in the country is the former “Žirovski Vrh” uranium mine, which is now in the final phase of restoration. There are also some NORM facilities, which are not legally considered as radiation facilities since their influence to the environment is supposed to be small and therefore out of the scope of regulation (thermal power plants, TiO2 production facility); Nevertheless they discharge certain amounts of natural radionuclides to the environment. Last but not least are hospitals with nuclear medicine departments which discharge short lived radionuclides into surface waters.
The aim of this paper is to present what radionuclides are discharged from the above mentioned facilities and which of these can actually be measured in the environment and reported to the competent authority. This comprehensive overview can be used to further refine operational monitoring programmes to better assess the impact on the environment.
_1102 ______
Task Analysis and Risk Assessment in Case of Accident Involving Transport of Radioactive Materials by Road in the Republic of Slovenia
Thomas BREZNIK, dipl. inž.rad, doc. dr. Marko Gerbec, doc. dr. Borut Smodiš;
Due to industrialization and simultaneous increasing heavy road traffic is the risk assessment of hazardous materials transport more and more important. Every year approximately a few hundred radioactive material packages (consumer goods excluded) are transported within, into and out of Slovenia by all modes of transport. Accidents and incidents related to transport operations in transport of radioactive materials – TRAM do and will inevitability occur from time to time. Therefore it is necessary to survey and evaluate all steps and aspects related to human error factor as main cause of
106 Nuclear Energy f0r New Eur0pe 2010 all emergency events in the transport chain so that we could take a proper remedial actions by preventing the reoccurrence of such events in the future. Human error can be considered as the main cause of all events related to TRAM.
Task analysis - TA is a fundamental methodology in assessment and reduction of human error. A wide variety of different task analysis methods exists and it would be impracticable to use all of them as evaluation tool. Therefore we extract and use only two of them. In general Hierarchical Task Analysis – HTA and Critical Action and Decision Evaluation technique – CADET can be used jointly as a framework for carrying out both action and cognitive task analysis.
Clearly an important element of risk assessment, human factors risk analysis is a continually growing field of interest in overall risk reduction strategies. There are six steps in performing a human factors risk analysis which display effective risk reduction strategy: 1. Break down the TRAM process into discrete tasks. 2. Identify human errors that may lead to system failure for each task. 3. Determine the worst-case effect of each error on the TRAM system. 4. Define barriers and controls to either prevent the error from occurring or mitigate its effect. 5. Perform a risk assessment of errors and their effect to determine how risks are to be addressed.
In order to perform such effective TA we will use a special task analysis tool (TaskAnalysis v2TM program) relied upon by many different organizations around the world for their mission critical projects such as radioactive materials transport operations. This article is therefore an account of TA work undertaken in TRAM with the purpose to:
• Promote and strengthen the system safety in TRAM in line with the latest scientific and technological developments in order to assess and reduce of human error occurrence. • Further development of emergency preparedness and response of radioactive material transport practice at all levels of transport chain • Ad final, to develop harmonised systems to promote transparency in the provision of information to improve the public perception of transport safety.
_1103 ______
Comparison of Measured Activity Concentrations in Plants from the Area of the Former Uranium Mine at Žirovski Vrh, Slovenia with Activity Concentrations Obtained by the ERICA Biota Assessment Tool
Černe Marko, Smodiš Borut “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
Uranium mining and milling activities are one of the major causes of radioactive contamination of the environment. Radionuclides, especially uranium decay products are discharged with U-mill tailings into the soil and water and consequently into vegetation where they accumulate. Transfer of radionuclides thus represent a radiological risk to humans and non-human biota due to accumulation of radionuclides in target tissues and ionising radiation. Uranium mine at Žirovski vrh in Slovenia, which operated from 1985-1990, processed about 600,000 tons of U-ore. U-milling and mining tailings were deposited at the Boršt and Jazbec sites. Risk assessment in such area could also be
107 Nuclear Energy f0r New Eur0pe 2010 evaluated using modelling approach as reported in recent studies. The ERICA assessment tool is one of more widely used modelling approach, that was developed to access the radiological risk to biota. ERICA tool is a software programme that guides the user through the assessment process and performs the necessary calculations to estimate risk to selected animals and plants. In the present study, the activity concentrations of 238U, 226Ra and 230Th in plants, such us soft rush (Juncus effusus L.), marsh marigold (Caltha palustris L.), tall moor grass (Molinia arundinacea (L.) Moench) and wood club-rush (Scirpus sylvaticus L.) determined by alpha-particle spectrometry were compared with activity concentrations predicted by ERICA –Biota Assessment Tool.
_1104 ______
A Brief History of Krško NPP Radiation Impact on Environment
Matjaž Koželj “Jožef Stefan” Institute Jamova 39, SI-1000 Ljubljana, Slovenia [email protected]
One of the most important requirements from the Site Permit of Krško NPP was related to the maximum annual effective dose on the 500 m border of the Power Plant protected area and beyond, which must not exceed 50 µSv per year. This request has never been relaxed and is still valid. In addition, Operational Licence issued in 1984 also included request regarding the limiting annual and quarterly emissions of radionuclides in river Sava, the value of which has been changed a few years ago in relation to the extended fuel cycle. Nevertheless, the changes have not decreased the level of protection of general public.
Krško NPP must provide the proof that it is complying with these requirements. Therefore Krško NPP is obliged to collect the data about the emissions of radionuclides and submit it the regulatory body, and also to provide monitoring of the environment in the surroundings of the power plant. The monitoring includes external dose measurements, sample collection and evaluation, as well as total dose assessment for members of public based on collected data and analytical models. Objectiveness and validity of results has been ascertained with the involvement of independent and authorised organisations in the monitoring implementation.
Requirements for monitoring and emissions reporting for Krško NPP have been in force from the very beginning of operation. The programme of measurements, which has been verified and approved annually, has been based on generic (and extensive) programme in the relevant rules from 1986 and the new rules from 2007. The main difference between these rules is not in the content of monitoring programme itself, but introduction of the additional requirements related to the quality and reliability of measurements. Nevertheless, the credibility and correctness of monitoring results in the past were never questionable, therefore now we have comparable monitoring data for more than twenty-five years of Krško NPP operation at our disposal.
It is our aim to briefly present and compare the extent and main results of available monitoring data from the beginning of Krško NPP operation until now, and also to discuss the differences in the data evaluation approaches used in eighties, nineties and in the last decade. We will also try to discuss the consequences of Chernobyl accident for plant monitoring and also solutions used in the monitoring during post-accident years to get the realistic evaluation of Krško NPP operation consequences for population and environment.
108 Nuclear Energy f0r New Eur0pe 2010
Education, Public Relations and Regulatory Issues
109 Nuclear Energy f0r New Eur0pe 2010
_1201 ______
A Survey of the Contribution of Large Experimental Nuclear Facilities to Education in Europe
Michel Giot Université catholique de Louvain and SCK•CEN Boeretang 200, B-2400 Mol, Belgium [email protected]
In the framework of the ETKM working group of SNE TP (Sustainable Nuclear Energy Technological Platform) a survey was performed to identify- and to quantify the use of- the experimental facilities available for the nuclear academic education in the European Union, Norway and Switzerland. Data were collected from both the supply side – the operators of the facilities – and the demand side – the academics in charge of Master and Doctoral programmes. The facilities considered in the survey include the research reactors and the thermal-hydraulic facilities, as well as plant simulators.
From the survey it appears that only 25 out of the 53 presently operated research reactors are effectively used for laboratory sessions at BSc and MSc levels, and that only 8 reactors are operated for this purpose during more than 120 hours per year each. From the operators’ point of view, an increase of at least 50% in the number of accommodated students is considered as possible, going from 2750 to 4100 students. The research reactors are also used for the preparation of MSc and PhD theses. Both numbers of theses range in the order of 70 theses per year.
While almost all research reactors are ageing, most thermal-hydraulic facilities have been built recently, offering a diversity of experimental tools (water, liquid metals, severe accidents, etc.), and are used for the production of about 20 PhD and 20 MSc theses a year. According to the operators of the facilities the numbers of accommodated students could be multiplied by a factor of 3 without too much problem.
Except for a few of them, the 35 academic institutions which filled in the questionnaire are rather satisfied with the present offer. However, they require paying a special attention to the nuclear research reactors located at university premises which are more intensively used. They also suggest improving the mobility of students by increasing the number of mobility grants, by organising experimental short courses at the premises of the facilities, and by pooling the facilities for student projects. These efforts should be supported by an ad hoc data base where all relevant information on the use of experimental facilities could be found, and by a foundation offering and distributing the grants.
The paper includes a set of tables where detailed results of the survey can be found.
110 Nuclear Energy f0r New Eur0pe 2010
_1202 ______
Public Opinion about Nuclear Energy – Year 20010 Poll
Radko Istenič, Igor Jenčič "Jožef Stefan" Institute Milan Čopič Nuclear Training Centre Jamova 39 Ljubljana, Slovenia [email protected]
Public information became one of the important activities of the Nuclear Training Centre at the Jožef Stefan Institute in Ljubljana. The Information Centre was established in 1993 with the primary goal to inform the visitors about nuclear power and nuclear technology in general and about Krško Nuclear Power Plant. A program of lectures for youngsters is supplemented by a permanently evolving exhibition about nuclear power and demonstrations of radioactivity phenomena.
The visit consists of a live lecture followed by the demonstration of radioactivity, a guided tour of exhibition and sometimes by a tour of the TRIGA research reactor.
Our main target group are schoolchildren with their teachers. Most of them are from the 8th and 9th grade of elementary school, age 14 to 15. Every year some 8000 youngsters visit the Information Centre.
We monitor the opinion trends since 1993 by polling about 1000 youngsters every year. The youngsters are polled before they listen to the lecture or visit the exhibition. In that way we are trying to obtain their unbiased opinion based on the knowledge from everyday life. In the paper we will present and comment the results of the 2010 poll. The trends that we monitor since 1993 are more indicative than the absolute numbers.
_1203 ______
Adopting Curriculum Integration to Improve Nuclear Training and Education
David Helling, David Kwiatkowski, Nathan Hall, Pamela Aigner Westinghouse Electric Company, LLC P.O. Box 159, Madison, Pennsylvania, U.S.A. [email protected], [email protected], [email protected], [email protected]
As the nuclear community expands and its population diversifies, so too must educational methodologies evolve in order to provide superior student learning. Westinghouse will use the curriculum integration approach to drive this evolution. Curriculum integration (CI) is an educational paradigm that brings all the aspects of the curriculum together into meaningful association to create broad areas of student competency. At its core, CI recognizes the necessity for students to see the big picture rather than requiring that learning be divided into small pieces.
CI is an important aspect of learning that allows students to see relationships among theory, technical concepts, and the application of these theories and concepts to plant operations. Students learn to engage in authentic interdisciplinary discussions that result in a deep understanding of the topic.
111 Nuclear Energy f0r New Eur0pe 2010
Consequently, students become more responsible for, and engaged in, their own learning as the instructor assumes more the role of a facilitator rather than simply an information dispenser. Curriculum integration permits the learning process to be more flexible and responsive to the various needs of students. When using CI, cultural background, age, and differing levels of experience become enhancements to learning rather than barriers.
Although new to the nuclear industry, curriculum integration is a well accepted methodology that is increasingly utilized by corporations and universities. Unfortunately, however, our industry largely continues to favor traditional, instructor centered methods that often fall short of developing the depth of understanding that is achieveable using CI. It is imparitive that the nuclear industry now acknowledge the inherent potential of its richly diverse population and seak the very best educational strategies available.
For the past ten years, Westinghouse has worked to improve interactive training techniques. Central to our success is the use of a progressive student-centered approach that establishes a learning partnership between and among the student, his or her management, and our instructors. Student- centered learning is now viewed as a subset of the more complete curriculum integration approach. In this presentation, we will discuss how and why a focus on curriculum integration with its emphasis on interactive student learning is being implemented and we will describe the performance improvement that has already been observed in the U.S. plants using these methods.
The learning approach that we will outline during this oral presentation leads directly to significant improvement in the skills and knowledge required to drive plant performance to increasingly higher levels. This is a learning approach that can significantly contribute to safe and vital Nuclear Energy for New Europe and for the world.
_1204 ______
Implementation of the Code of Conduct on the Safety and Security of Radioactive Sources in Slovenia
Tatjana Frelih Kovačič, Janez Češarek, Igor Osojnik, Maksimilijan Pečnik Slovenian Nuclear Safety Administration Železna cesta 16, p.p. 5759, SI-1001 Ljubljana, Slovenia [email protected], [email protected]
The revised Code of Conduct on the Safety and Security of Radioactive Sources (CoC) was published by the International Atomic Energy Agency (IAEA) in January 2004. It has been strengthened to take account of international concerns following the events of September 11. Although the CoC is not a legally binding instrument, on February 3, 2010, there were 97 states expressed their full support and endorsement of the Agency's efforts to enhance the safety and security of radioactive sources. It provides guidance on how states can safely and securely manage radioactive sources that may pose a significant risk. As a supplementary guidance to the CoC the Guidance on the Import and Export of Radioactive Sources was subsequently published by the IAEA in March 2005.
In Slovenia, the current system of regulatory control of radioactive sources assures the compliance with the objectives of the CoC and of the Guidance on the Import and Export of Radioactive Sources. The paper is presenting the current circumstances in the country with regard to the implementation of CoC, highlighting some future tasks and challenges as well. The measures necessary to ensure that the radioactive sources within Slovenian territory are safely managed and securely protected during their life-time are described, e.g. appropriate legislation and regulations, establishment of the regulatory
112 Nuclear Energy f0r New Eur0pe 2010 bodies, establishment of a national register of a radioactive sources, appropriate national strategy for control over orphan sources, appropriate facilities and services available to the persons authorised to manage radioactive sources.
_1205 ______
Training of Specialists in the Assessment of NPP Equipment Compliance
Evgeny Kapralov, Yury Kapralov, Gennady Filimonov Training and Methodological Center for Nuclear and Radiation Safety (TMC NRS) 8 Parkovaya, 9, 105523, Moscow, Russia [email protected]
The Russian Federation is now developing an ambitious state program for NPP construction. With the two units being constructed annually, serious attention shall be paid to the supervision over the quality of manufacture of the NPP equipment and assessment of its compliance with the effective safety codes and standards in atomic energy use.
There is a state regulatory document (NP-071-06) in the nuclear sector of the Russian Federation, which defines the acceptance performed by the competent expert organizations on behalf of the operating company as one of the forms of compliance assessment of the NPP equipment, materials and semi-finished products.
However, with the existing volume of the NPP equipment manufacture and supply to fulfill the objectives set by the state program for NPP construction, and large number of equipment manufacturers and suppliers, the specialized expert organizations experience an acute shortage of manpower, acceptance inspectors, in particular.
To address the challenge, based on the competence analysis made by the professional expert community and in close collaboration with the major expert organization for equipment acceptance VO “Safety”, the TMC NRS has elaborated a refresher training course “Supervision over the quality of NPP equipment manufacture and acceptance inspections”.
The training course has the following training modules: principles of nuclear and radiation safety assurance; basic types of NPPs and their equipment; lifecycle of NPP products; measurement assurance of equipment manufacture; basic types of manufacturing documentation and procedure of its approval; arrangement of plant quality control; arrangement and performance of compliance assessment as a technical acceptance; basic methods and quality control, basic inspection/testing methods.
The training materials developed for the training course can be used both for classroom instruction and distant education of the trainees. The training materials are now being adapted to enable e-learning training.
More than 100 specialists were trained under the training course in 2009. Training and methodological manuals were developed and published in the National Research Nuclear University (NRNU) MEPhI. The manuals are now being translated into English.
To streamline the training in the acceptance of the NPP equipment, a pilot project for a “throughout” education (college-university-specialized training center) has been developed. The project will allow one to elaborate a new educational standard for higher professional training in the NPP equipment
113 Nuclear Energy f0r New Eur0pe 2010 assessment, which will enhance the training quality, ensure the sufficient number of manpower, thereby meeting the challenge posed by the Russian nuclear sector.
_1206 ______
Analysis of Human Resources and Technical Knowledge needed for the Licensing of the New Nuclear Build: the SNSA Approach
Siniša Cimeša, Andreja Peršič, Leopold Vrankar, Andrej Stritar Slovenian Nuclear Safety Administration [email protected]
Slovenia is seriously considering building a new nuclear power plant (NPP). For the Slovenian Nuclear Safety Administration (SNSA) it is necessary to undergo some basic preparations. At the moment the SNSA does not have sufficient resources for licensing and overseeing the design, construction and operation of the possible new plant. Likewise, the question arises whether technical support organizations which support the Administration in supervising the existing Krško NPP have sufficient capacity.
At the beginning of 2009 the SNSA’ project team prepared the analysis of licensing process, which is basically an overview of spatial planning, construction and nuclear safety regulation processes. One of the main results of this analysis was the list of tasks that would need to be performed by the SNSA in the new NPP licensing process. Regarding those results and taking into account the number of current SNSA employees, the next challenge is to prepare good arguments and strategy for new recruitments to elude current policy of decreasing public workers and to assure the adequate number of well trained staff, which are crucial for the efficient and timely execution of all tasks of the licensing process. This will enable the SNSA to establish a qualified and effective infrastructure for a possible new nuclear build.
_1207 ______
Deployment of New Nuclear Power Plant in Slovenia
Tea Bilic Zabric INKO consulting Kolezijska 5a, SI-1000 Ljubljana, Slovenia [email protected]
The licensing process for construction of a nuclear facility is divided in two stages. The first stage is for implementing a spatial planning process to allocate and approve a suitable area or site based on a plan of the project. The second stage is for issuing of construction license.
Changes in the regulatory regimes were throughout the world due to the increased attention to nuclear safety and environmental protection, as well as the advanced technologies and the appeared new design solutions. Nowadays, in many countries the site selection for and the construction of a nuclear power plant is not possible without the positive consent of the population most directly affected.
The second nuclear power unit is planned to be constructed near the Krsko NPP.
Even it looks that spatial planning is well defined in the Slovenian legislation (this area is covered by a variety of regulations, from nuclear to environmental and construction), it is a highly complex and
114 Nuclear Energy f0r New Eur0pe 2010 challenging process, with several questionable areas. What makes it even more complex are multiple interactions and relationships between elements of the processes, but also the actors involved.
Regulatory status and arrangements regarding spatial planning of new NPP unit in Slovenia and comparison of Slovenian arrangements and practices with requirements, practices and arrangements of selected EU countries, that are preparing for or already committed to build new nuclear power plants, is a subject of this article.
_1208 ______
EHRO-N: a Tool Complementing Instruments and Initiatives for Improved Management of Nuclear Human Resources in the European Union
Veronika Simonovska, Ulrik Von Estorff “Joint Research Centre – Institute for Energy” P.O. Box 2, NL-1755 ZG Petten, Netherlands [email protected], [email protected]
Advances in science and technology require higher quantity and better quality of human resources than are available today in the European Union (EU) in all three spheres covered by the knowledge triangle: education, innovation, and Research and Development. The situation concerning nuclear human resources has deteriorated even further in the EU in the past decades and there is a risk of the loss of important nuclear knowledge if no actions are taken. This prompted the Council of the EU to conclude in December 1st, 2008, that it is “essential to maintain in the European Union a high level of training in the nuclear field” and, at the same time, preserve the skills in the nuclear field that we already have. The latter is especially true if we are to guarantee the highest level of nuclear safety (subject of the Council Nuclear Safety Directive adopted in June 25th 2009).
The nuclear knowledge management landscape in the EU includes numerous instruments and initiatives tackling the nuclear human resources challenge. The European Commission (EC), more precisely the Directorate-General (DG) for Research with its initiatives Sustainable Nuclear Energy Technology Platform (SNE-TP) and the European Nuclear Education Network (ENEN) and DG Energy with its initiative the European Nuclear Energy Forum (ENEF) are at the forefront of these efforts. In ENEF was born the idea of European Human Resources Observatory for the Nuclear Energy Sector (EHRO-N).
By creating a database of nuclear skills needed in the short-, medium-, and long-term perspective and by identifying gaps and deficiencies in the educational and training infrastructure in the EU, EHRO-N should link supply and demand for nuclear human resources in the EU. In cooperation with EC, ENEN, ENEF, SNE-TP, EHRO-N will work in favour of the development of a European scheme of nuclear qualifications and mutual recognition. Cooperation and coordination with international actors such as IAEA is envisaged since the situation regarding nuclear human resources is similar throughout the world.
115 Nuclear Energy f0r New Eur0pe 2010
_1209 ______
Krško NPP Full Scope Simulator Utilization Experience
Igor Fifnja, Matjaž Žvar, Dušan Češnjevar Nuklearna elektrarna Krško Vrbina 12, 8270 Krško, Slovenia [email protected], [email protected], [email protected]
During the past three decades, control room simulators became very important tool used for training of nuclear power plant personnel. There are various international organizations (IAEA, WANO…) that provide exchange of information and experience as well as documents providing guidance on simulator utilization and configuration management.
At Krško NPP, we have started to use our own full scope simulator in April 2010, as a part of major plant modernization project involving replacement of Steam Generators and plant power uprate. In the past, our operators used to be trained on various simulators in United States while there was a long- lasting quest for our own, plant specific simulator. The simulator project, consisting of preparation phase, simulator design, construction, testing and installation on-site, took almost five years.
Since then Krško NPP full scope simulator was to large extent used to support following two major operations personnel training programs: - Initial licensed operator training; the program used to provide comprehensive training for future main control room operators. Program in total duration of 85 weeks consists of following four phases: Fundamentals, Plant systems and operation, Simulator training and On-the-job training. During this program simulator is mainly used during second and third phase. - Licensed operator continuing training; the program used to provide refresher training for already qualified operators. This training is delivered in two-year cycles, comprised of four training segments per year and including up to 80 hours of simulator training for each operating crew annually. In addition, this program includes training on simulated virtual local panels during non-licensed operators training and running of certain number of scenarios for entire operating crews.
There are other training programs and plant activities where Krško NPP simulator was effectively utilized during this time period: Emergency Preparedness Organization drills, training of other plant or external personnel, development and validation of operating procedures, testing of plant modifications, etc.
Simulator Configuration Management Program and supporting processes were effectively established and organized to ensure that simulator is kept consistent with plant status and to enable quality training of operators. The configuration management processes involve screening and implementation of plant modifications that have influence on the main control room operation or simulator response, conduct of regular simulator testing to demonstrate simulator correct response and fulfill international standards requirements, continuous maintenance of simulator, etc.
During the past 10 years the Krško NPP full scope simulator was introduced in different plant processes, especially training of operations personnel. There are many indicators demonstrating that the simulator was effectively utilized and maintained which provides a good foundation for the future.
116 Nuclear Energy f0r New Eur0pe 2010
_1210 ______
Education and Training of Future Nuclear Engineers at DIN: From Advanced Computer Codes to Interactive Plant Simulator
O. Cabellos, C.Ahnert, D.Cuervo, N.García-Herranz, E.Gallego, E.Mínguez, J.M.Aragonés, A.Lorente, D.Piedra Departamento de Ingeniería Nuclear (DIN), Universidad Politécnica de Madrid José Gutiérrez Abascal 2, 28006 Madrid, Spain corresponding.author: [email protected]
This paper presents the work performed at the Department of Nuclear Engineering (DIN) of the Universidad Politécnica de Madrid to improve the education and training of future Spanish nuclear engineers: (i) The experience gained in the last years by our Department in the simulation of the Nuclear Power Plants, mainly in PWR, has been included in the “Nuclear Technology” programme with different subjects about optimization of fuel reloading, manoeuvres, start-up, …
(ii) In addition, the “Nuclear Reactor Design” programme has been focused on the understanding of the advanced computational codes for nuclear reactor designs, starting with the nuclear data processing codes, then the core calculations codes, and finally the plant simulators codes (JANIS, NJOY, WIMSD, ORIGEN/ACAB, MCNP, COBAYA/SIMULA, COBRA, SIMTRAN, RELAP). Some of these codes have been developed in our Department for many years.
(iii) But more realistic studies are also required to complete the education&training objectives in the “Nuclear Safety” programme, and in this sense the use of a simulator is the appropriate tool to be used. The International Atomic Energy Agency (IAEA) sponsors the development of nuclear reactor simulators for education, or arranges the supply of such simulation programs. Aware of this, the DIN was provided in 2008 with the Interactive Graphical Simulator of the Spanish nuclear power plant José Cabrera, whose operation ceased definitively in 2006. The simulator is a Graphical Simulator, used for training of main control room personnel, technical support engineers, and operations management. This paper presents the work performed at the Department to turn the simulator into a teaching/learning tool, to be use in the nuclear engineering studies.
117 Nuclear Energy f0r New Eur0pe 2010
_Index of Authors D Helling, David 111 D'Auria, Francesco 21, 48, 49, Henry, François 35 A 76 Hermansky, Peter 47 Adorni, Martina 49 Debrecin, Nenad 68 Herranz, Luis E. 71 Agresta, Giuseppe 48 Degmova, Jarmila 54 Hollerbach, Mark 37 Ahnert, C. 117 Dekan, Julius 54 Hortal, Javier 56 Aigner, Pamela 111 Del Nevo, Alessandro 49, 76 Amižić, Milan 68 Dessars, Nathalie 67 I Anghel, Ionut 34 Destouche, Christophe 21 Ibáñez, Luisa 56 Anglart, Henryk 34 Dimovic, Slavko 83 Iliev, Atanas 63 Antola, Lorenzo 11 Dominguez, Cristina 70 Istenič, Radko 111 Aragones, J.M. 117 Dossantos-Uzarralde, P. 17 Ivekovič, Aljaž 102, 103 Araneo, Dino 48 Draksler, Martin 87 Iveković, Ilijana 41 Avsec, Jurij 7 Dražić, Goran 91, 102, 103 Izquierdo, Jesus 94 Duhovnik, Boštjan 80 Izquierdo, Jose M. 56 B Duhovnik, Jože 95, 101 Babazadeh, Davood 61 Duponcheel, Matthieu 36 J Bajs, Tomislav 41 Jacquet, H.P. 17 Balat - Pichelin, Marianne 89 E Jaekel, M. 16 Barrachina, T. 46, 51 Estrada, Herb 30 Ječmenica, Radomir 16 Bartosiewitz, Yann 35, 36 Estrada-Perez, Carlos 43 Jelić, Nikola 95, 101 Bašić, Ivica 41 Evstyukhin, Nikolay 57, 59 Jenčič, Igor 111 Bedogni, Roberto 99 Jordan, Thomas 68 Benčik, Vesna 42 F Juanas, J. 51 Bergholz, Steffen 58 Fabjan, Marija 81 Beuzet, Emilie 75 Faramarzi, Jay 37 K Beyer, Matthias 31 Fernandez, Antonio 62 Kačegavičius, Tomas 42 Bilić-Zabric, Tea 114 Fernández, Ivan 56 Kadivnik, Dejvi 26 Böck, Helmuth 17 Fifnja, Igor 116 Kallemets, Kalev 13 Božič, Simon 58 Filimonov, Gennady 113 Kančev, Duško 61 Breznik, Thomas 106 Finnicum, David 73 Kapralov, Evgeny 59, 113 Bricteux, Laurent 36 Fiorina, Carlo 7, 9 Kapralov, Yury 57, 59, 113 Bubeck, Dave 25 Forrest, Robin 2 Karimi, Kaveh 38 Bukošek, Vilibald 91 Frelih Kovačič, Tatjana 112 Keresztúri, András 50 Bundara, Borut 58 Khan, Rustam 17 G Kljenak, Ivo 32 C Gaio, Paolo 8 Kobe, Spomenka 91 Cabellos de Francisco, Oscar Gallego, E. 117 Končar, Boštjan 43, 87 117 Garcia Herranz, N. 117 Kos, Leon 101 Calabrese, Rolando 14 Garrido, Oriol 54 Kostadinov, Venceslav 72 Cambriani, Andrea 11 Gauriot, G. 17 Kostanjevec, Marko 81 Cammi, Antonio 7, 9, 39, 40 Gaus-Liu, Xiaoyang 68 Kotev, G. 21 Canamón, Israel 56 Gaveau, B. 16 Kovačič, Jernej 93, 99 Catsaros, Nikolas 16 Gerbec, Marko 106 Koželj, Matjaž 108 Chambers, Martin 26 Ghasemizad, Abbas 100 Krajčovič, Marian 47 Churbanov, Aleksander 33 Gil, Jesus 56 Krajina, Dražen 25 Cimeša, Siniša 114 Giot, Michel 110 Kralj, Metka 84 Cindro, Michel 106 Gjorgiev, Blaže 63 Krek, Janez 95 Cizelj, Leon 53, 54, 55 Gómez, Javier 56 Krepper, Eckhard 31 Colombo, Marco 40 Grabnar, Zvone 96 Križman, Milko 106 Conroy, Sean 88 Grande, L. 12 Kromar, Marjan 18, 25 Coscarelli, Eugenio 76 Grgić, Davor 16, 42 Kropík, Martin 27 Cuervo, D. 117 Grlj, Nataša 96 Kurinčič, Bojan 18, 26 Cvelbar, Robert 58 Guerrieri, Claudia 7, 9 Kwiatkowski, David 111 Gyergyek, Tomaž 93, 99 Č L Čadež, Iztok 96, 104 H Lamy, Jean 75 Čepin, Marko 61, 63, 64 Hall, Nathan 111 Lengar, Igor 88 Čerček, Milan 93, 99 Hassan, Yassin 43 Lenošek, Melita 90 Černe, Marko 107 Haste, Tim 70 Leskovar, Matjaž 72, 76 Češarek, Janez 112 Hauser, Ernest 30 Levanat, Ivica 84 Češnjevar, Dušan 116 Hedberg, Stellan 34 Lipka, Jozef 54 Heffron, Raphael 13 Lizon-A-Lugrin, Laure 10 Hegyi, György 50 Lokner, Vladimir 84 118 Nuclear Energy f0r New Eur0pe 2010
Lorente, A. 117 Poeckl, Christian 58 T Lucan, Dumitra 53 Popovichev, Sergey 88, 92 Tavčar, Polona 81 Lucas, Dirk 31 Porosnicu, C. 97 Teyssedou, Alberto 10 Lungu, C. 97 Predin, Andrej 7 Toplišek, Tea 91, 103 Lutz, R. 67 Prošek, Andrej 35 Tóth, Ignac 54 Luzzi, Lelio 7, 9 Trenta, Fabrizio 11 Q Trkov, Andrej 21, 99 M Queral, César 56 Tromm, W. 68 Macián, R. 51 Trosztel, Istvan 50 Maillard, Jacques 16 R Maŕaczy, Csaba 50 Rantamäki, Reko 4 U March, Philippe 70 Rapić, Andrea 84 Uplaznik, Mihaela 55 Markelj, Sabina 96, 104 Rataj, Jan 27 Urbančič, Andreja 6 Martínez-Murillo, J.C. 46 Rensonnet, Thibaut 67 Uršič, Mitja 72, 76 Massaut, Vincent 98 Ribič, Mirko 96 Uytdenhouwen, Inge 98 Matijević, Mario 19 Ricotti, Marco 40 Maurel, G. 16 Rizzo, Paul 62 V Mavko, Borut 32, 76 Robledo, Fernando 71 Van Oost, Guido 98 Mele, Irena 84 Rozzia, Davide 49 Varvayanni, M. 16 Meléndez, Enrique 56 Rudolph, Juergen 58 Vasiliev, Alexander 77 Melikhov, Oleg 40 Rupnik, Zdravko 96, 104 Vavpetič, Primož 96 Melikhov, Vladimir 40 Verdú, G. 46, 51 Meyer, L. 68 S Vesel, Alenka 89, 89 Miassoedov, Alexei 68 Saltanov, Eu. 12 Villalba, Ernesto 56 Miletić, M. 12 Sanchez, Miguel 56 Virtič, Peter 7 Miller, Herbert 37 Sartori, Enrico 2 Visweswaran, Srinivasa 73 Milocco, Alberto 99 Savva, P. 16 Vleugels, Jozef 98 Minguez, E. 117 Schimann, Peter 24 Vojnovič, Djordje 56 Miró, R. 46, 51 Schmidtke, Martin 31 Vokal Nemec, Barbara 106 Mokry, Sarah 12 Sheng, Hua 98 Volkanovski, Andrija 64, 64 Moosavi, M. 100 Silva, J. 16 Von Estorff, Ulrik 115 Morozov, Anton 57 Simondi-Teisseire, B. 70 Vragolov, Dinka 19 Mozetič, Miran 89, 89 Simoni, Eric 75 Vrankar, Leopold 114 Muehleisen, Artur 56 Simonovska, Veronika 115 Murcia, Santiago 56 Simonovski, Igor 53, 54, 55 W Slavcheva, Katia 11 Wirtz, Nikolaus 58 N Slavič, Slavko 25 Nemanič, Vincenc 97 Slugeň, Vladimír 54 X Nerovnov, Alexey 40 Smieško, Ivan 54 Xu, Cheng 4 Novak, Saša 90, 102, 103 Smodiš, Borut 106, 107 Smutek, Jan 83 Y O Snoj, Luka 18, 88, 92 Younesian, Elham 20 Osojnik, Igor 81, 112 Solovjanov, Oleg 67 Yousefpour, Faramarz 38, 61 Soltani, Hamid 38, 61 P Spinelli, Benedetto 9 Z Papini, Davide 39, 40 Stanič, Ana 82 Zajec, Andrej 58 Parfenov, Yuriy 40 Steinbrück, Martin 68, 70 Zajec, Bojan 97 Parisi, Carlo 21 Stošić, Zoran V. 24 Zaplotnik, Rok 89, 89 Pastore, Giovanni 39 Stritar, Andrej 56, 114 Zeyen, Roland 70 Payot, Frédéric 70 Stropnik, Primož 84 Zieger, Tobias 37 Pazirandeh, Ali 20 Suban, Marjan 58 Pecchia, Marco 21 Sučič, Boris 6 Ž Pečnik, Maks 81, 112 Sučić, Simona 81 Žagar, Tomaž 7 Peiman, W. 12 Surin, Vitaly 57, 59 Žefran, Bojan 25 Pelicon, Primož 96 Svoboda, Jiří 83 Železnik, Nadja 79, 84 Pereira, C. 46 Syme, Brian 3, 88, 92 Žerovnik, Gašper 18, 21 Perron, Hadrien 75 Žigman, Vida 94 Peršič, Andreja 56, 114 Š Žvar, Matjaž 116 Pevec, Dubravko 16, 19 Šadek, Siniša 42, 68 Piedra, A. 117 Šeliga, Pavol 54 Pillon, M. 99 Špiler, Janja 80 Pioro, Igor 10, 12 Štrubelj, Luka 7 Plećaš, Ilija 83 Podjavoršek, Matjaž 56 119