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Book of Abstracts

24th International ConferenceNuclear Energy for New Europe

Portorož, September 14 ‒ 17, 2015www.nss.si/nene2015/

V O D A + S O N C E + J E D R S K A E N E R G I J A

E N E R G I J O N A R A V E P R E V A J A M O V E L E K T R I K O .

Valovanje je izmenjava energije med delci snovi. Gibanje vode, svetloba, toplota so valovanja, trajnostni viri energije, ki omogoËajo življenje. V skupini GEN ta valovanja zanesljivo, varno in okolju prijazno spreminjamo v elektriËno energijo, s katero oskrbujemo porabnike.

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Book of Abstracts

24th International Conference Nuclear Energy for New Europe i

Organizers

Nuclear Society of Slovenia

www.nss.si/nene2015/

Contact AddressNuclear Society of Slovenia

NENE2015Jamova cesta 39SI-1000 Ljubljana

Slovenia

T +386 1 588 53 02

F +386 1 588 53 76

E [email protected]

Jožef Stefan Institute, Nuclear Training Centre (ICJT)

Book of Abstracts

ii 24th International Conference Nuclear Energy for New Europe

Committees

Program committee

Igor Jenčič, Slovenia

Christophe Béhar, FranceHelmuth Böck, AustriaLeon Cizelj, Slovenia

Gerard Cognet, FranceNikola Čavlina, CroatiaMarko Čepin, SloveniaFrancesco D’Auria, ItalyMilorad Dusic, AustriaMichel Giot, Belgium

Miroslav Gregorič, SloveniaDavor Grgić, Croatia

Pavlin Petkov Groudev, BulgariaTomaž Gyergyek, Slovenia

Tim Haste, FranceTomasz Jackowski, Poland

John E. Kelly, USAIvan Kodeli, Slovenia

Boštjan Končar, SloveniaZdenek Križ, Czech Republic

Matjaž Leskovar, SloveniaBorut Mavko, Slovenia

Irena Mele, IAEAPrimož Pelicon, SloveniaDubravko Pevec, CroatiaStane Rožman, SloveniaRainer Salomaa, Finland

Igor Simonovski, ECLuka Snoj, Slovenia

Andrej Stritar, SloveniaIztok Tiselj, SloveniaAndrej Trkov, IAEA

Eugenijus Ušpuras, LithuaniaSimon Walker, United Kingdom

Tomaž Žagar, SloveniaNadja Železnik, Slovenia

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24th International Conference Nuclear Energy for New Europe iii

Organizing committee

Mateja Južnik, Slovenia

Saša Bobič, SloveniaJure Hribar, Slovenia

Borut Mavec, SloveniaVesna Slapar Borišek, Slovenia

Luka Tavčar, SloveniaNina Udir, Slovenia

Bojan Žefran, Slovenia

Young author award committee

Iztok Tiselj, Slovenia

Michel Giot, BelgiumTim Haste, France

Dubravko Pevec, Croatia

Best poster award committee

Gerard Cognet, France

Helmuth Böck, AustriaPrimož Pelicon, Slovenia

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iv 24th International Conference Nuclear Energy for New Europe

Previous meetings organised by the 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 2005, Bled, Slovenia, September 2005






• Nuclear Energy for New Europe 2011, Bovec, Slovenia, September 2011

• Nuclear Energy for New Europe 2012, Ljubljana, Slovenia, September 2012



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24th International Conference Nuclear Energy for New Europe v

WelcomeNuclear Society of Slovenia welcomes you at the traditional meeting of professionals from nuclear research organizations, educational institutions, nuclear utilities, industrial companies and regulatory bodies. This meeting has a long tradition as it has evolved from a national conference to a regional meeting and after a decade it has gained a true international character.

The 2015 conference’s invited lectures span from comparative studies of different electricity supply systems to solving practical NPP operational problems using advanced scientific methods, as well as results and future directions of fusion research. Contributed papers cover a wide range of current developments in different fields related to nuclear industry, research, education and regulation. The opportunities and challenges for nuclear power generation will be highlighted and discussed. Professionals who recognize nuclear power’s importance in securing Europe’s energy and environmental future are very welcomed to attend this annual conference.

Place and time of conferenceThe conference will take place in Grand Hotel Bernardin in Portorož, Slovenia. GH Bernardin is the first and the largest convention hotel in Slovenia. Its wonderful location close to the sea provides inspiring working atmosphere.

St. Bernardin ResortGrand Hotel BernardinObala 26320 Portorož

Lectures and poster sessions will be held in the Europa Convention Halls “A” and “C“ on the 12th

floor of Grand Hotel Bernardin.

From: Monday, September 14, at 16:00

To: Thursday, September 17, at 14:00

St. Bernardin Resort Grand Hotel Bernardin

Hotel Vile ParkHotel Histrion

Piran

Portorož

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vi 24th International Conference Nuclear Energy for New Europe

PresentationsOral presentations are limited to 20 minutes. Authors are kindly asked to present their papers in 15 minutes to allow 5 minutes for discussion. A laptop PC with Microsoft PowerPoint will be available.

Participants are asked to deliver their presentation files (CD, memory stick, ...) to the organizer at least one break before their scheduled presentation time. For video projectors or any additional questions please contact the organizers.

Posters should fit within the 95 cm (width) x 110 cm (height). Authors are kindly requested to post their presentations on Tuesday, September 15 morning and remove them by Thursday, September 17, 13:00. Poster sessions with authors present at their posters are scheduled for Tuesday, September 15 at 15:40 and for Wednesday, September 16 at 10:10.

The best poster presentation will receive a special award.

ProceedingsThe Proceedings containing full-length papers presented at the conference, and accepted after peer review, will be published on the DVD after the conference and sent to the participants.

Contest of Young AuthorsAward will be given for the best paper prepared by the first author aged no more than 32 years in 2015. Full papers should have been submitted by August 31, 2015 in order to enter the contest.

Best Poster AwardPosters will be evaluated according to their clarity, technical content and visual attractiveness. Posters must be displayed from Tuesday, September 15, 2015 in order to participate in the contest.

RegistrationRegistration desk opening hours:Monday, September 14: 15:00 to 19:00

Tuesday, September 15: 8:00 to 18:00

Wednesday, September 16: 8:00 to 12:30

Thursday, September 17: 8:00 to 12:00

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24th International Conference Nuclear Energy for New Europe vii

Social Activities

Conference Lunch Tuesday, Wednesday and ThursdayLunch on Tuesday, Wednesday and Thursday is included in the registration fee and will be served from 12:30 to 13:30 at Grand Restaurant at the 10th floor of the Grand Hotel Bernardin.

Welcome Reception Monday, September 14The Welcome Reception will start at 19:30 in Grand Garden terrace at 11th floor of the Grand Hotel Bernardin with welcome drink.

Conference Trip Wednesday, September 16An afternoon trip to the Škocjan Caves will be organized on Wednesday afternoon.

Conference Dinner Wednesday, September 16The Conference dinner will be another opportunity for relaxed discussions. It will be held at 19:30 on the Terrace International, located by the square with the church, between hotels Bernardin and Histrion. In case of bad weather the dinner will take place in the Sunset Restaurant on the 10th floor of the Grand Hotel Bernardin. The Young Authors Award and the Best Poster Award will be presented during the dinner.

the Terrace InternationalHotel Vile Park

Hotel Histrion

All social activities are free of charge for registered conference participants.

Program

viii 24th International Conference Nuclear Energy for New Europe

Conference TimetableAll lectures and poster sessions will be held in Grand Hotel Bernardin at Europa Convention Halls “A” and “C“.

Monday, Sept. 14 Tuesday, Sept. 15 Wednesday, Sept. 16 Thursday, Sept. 17

Invited Lecture Simon P. Walker

8:30 - 9:10 500 Severe Accidents &Probabilistic Safety

Assessment 8:30 - 10:10

Invited Lecture Karl Krieger 8:30 - 9:10

200 Thermal Hydraulics I

9:10 - 10:10

700 Nuclear Fusion

9:10 - 10:10

Coffee breakPosters with Coffee break 10:10 - 11:10

Coffee break

300 Research Reactors

10:30 - 11:30800

Materials 10:30 - 12:30200

Thermal Hydraulics II 11:10 - 12:30

600 Radioactive Waste,

Environmental Issues 11:30 - 12:30

Lunch Lunch Lunch

400 Reactor Physics I

14:00 - 15:40

Conference trip14:00 - 19:00

Posters with Coffee break 15:40 - 16:20Conference opening

16:00

Invited Lecture Stefan Hirschberg

16:20 - 17:00400

Reactor Physics II 16:20 - 18:00

Coffee break

100 Nuclear Energy &

Society 17:20 - 18:40

Welcome reception 19:30 -

Conference dinner 19:30 -

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Conference Program

Monday, September 14

16:00 Conference Opening

Invited lecture Chairperson: IgorJenčič,Slovenia

16:20 1 Stefan Hirschberg-Switzerland Sustainable Electricity: Wishful thinking or near-term reality?

Session 1 17:20 Nuclear Energy and Society Chairpersons: StanislavRožman,Slovenia

TimHaste,France

17:20 101 ChristopheBéhar,Pierre Le Coz-France Nuclear Energy, Challenges for the Future

17:40 102 Anna Przybyszewska,KajetanRozycki-Poland The need of development Gas Cooled Reactor Technology in Europe

18:00 115 Nikola Popov-Macedonia Nuclear Power Opportunities Through Regional Cooperation

18:20 104 Nadja Železnik-Slovenia How people perceive ionizing radiation: comparison in four countries

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x 24th International Conference Nuclear Energy for New Europe

Tuesday, September 15

Invited Lecture

08:30 2 Simon P. Walker-UnitedKingdom The analysis of crud light water reactor fuel rods, and its thermal hydraulic consequences

Session 2 09:10 Thermal Hydraulics I Chairpersons: IztokTiselj,Slovenia

MichelGiot,Belgium

09:10 201 Maria-Jose Rebollo-Spain Application of ISA methodology to a Loss of Normal Feedwater ATWS with TRACE 5.0

09:30 202 Andrej Prošek,BoštjanKončar,MatjažLeskovar-Slovenia Uncertainty Quantification of NEPTUNE_CFD calculation by Optimal Statistical Estimator Method

09:50 213 Matej Tekavčič,BoštjanKončar,IvoKljenak-Slovenia Influence of Liquid Inlet Modeling on Simulated Wave Characteristics in Vertical Gas-Liquid Churn Flow

Session 3 10:30 Research Reactors Chairpersons: BorutSmodiš,Slovenia

HelmuthBöck,Austria

10:30 301 Helmuth Böck,MarioVilla,AndreaBorioDiTigliole,JudyVyshniauskas, LubomirSklenka,LukaSnoj,AttilaTormási-Austria Human Resource Development and Nuclear Education Through Research Reactors: Successful Approach to Build Up the Future Generation of Nuclear Professionals

10:50 302 Tanja Kaiba,GašperŽerovnik,LukaSnoj-Slovenia Measurements with Multiple In-core Fission Chambers at the JSI TRIGA Mark II Reactor

11:10 304 Romain Henry,MarkoMatkovič-Slovenia Analysis of coolant temperature distribution for the validation of TRIGA Mark II CFD computational model

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Session 4 11:30 Radioactive Waste, Environmental Issues Chairpersons: MatjažKoželj,Slovenia

TomažŽagar,Slovenia

11:30 601 Tomaž Žagar,LeonKegel-Slovenia Preparation of the national program for the spent fuel and radioactive waste management taking into account possibility of European Repository Development Organisation development

11:50 602 PrimožMlakar,BoštjanGrašič,Marija Zlata Božnar,BorutBreznik-Slovenia On-line relative air dispersion concentrations one week forecast for Krško NPP prepared for routine and emergency use

12:10 603 Petra Planinšek,BorutSmodiš,LjudmilaBenedik-Slovenia Transfer of Th-230 from soil contaminated with U-mill tailing to radish, savoy and rocket

Session 5 14:00 Reactor Physics I Chairpersons: MarjanKromar,Slovenia

GyörgyHegyi,Hungary

14:00 401 Fausto Franceschini,MarjanKromar,AndrewT.Godfrey-USA Simulation of the NPP Krško Core at Hot Full Power with CASL Core Simulator - VERA-CS

14:20 402 Marcin Bielewicz,ElżbietaStrugalska-Gola,StanisławKilim,MarcinSzuta -Poland Experimental study of the physical properties of ADS systems – measurement of high energy neutron fields by using the 89Y threshold detectors.

14:40 403 Jakub Lüley,StefanCerba,BranislavVrban,JánHaščík,VladimirNečas-Slovakia Sensitivity Analysis of Gas-cooled Fast Reactor

15:00 404 Stephane Bourganel,JacquesDi-Salvo,ThiollayNicolas,SoldevilaMichel-France Preliminary Analysis of The FLUOLE-2 Experiment

15:20 405 Branislav Vrban,StefanCerba,JakubLüley,JánHaščík,VladimirNečas-Slovakia ALLEGRO uncertainty and similarity evaluation

Session 6 15:40 Poster Session Posters Session will also be held on Wednesday at 10:10.

Chairpersons: GerardCognet,France

HelmuthBöck,Austria

PrimožPelicon,Slovenia

Nuclear Energy and Society 103 Nikola Popov-Macedonia Human Resources Infrastructure Requirements for New Nuclear Power Program

105 Igor Simonovski,BojanŽefran,SteveClements,SandiCimerman-Netherlands The Role of High Performance Computing in the Nuclear Energy Sector

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xii 24th International Conference Nuclear Energy for New Europe

106 Vesna Slapar Borišek-Slovenia Use of the Geiger-Müller counter and the cloud chamber to present properties of radioactivity to youngsters

107 Radko Istenič,IgorJenčič-Slovenia Public Opinion about Nuclear Energy – Year 2015 Poll

108 Francine Menzel,FrancescoD’Auria,GaianeSabundjian,AlziraMadeira-Brazil Proposal of a BEPU-FSAR

109 Nadja Železnik,KjellAndersson-Slovenia Inputs for national research strategies for coordination of social, societal and governance issues in nuclear energy

110 DanielaDiaconu,MarinConstantin,GeorgiosGlinatsis,GiacomoGrasso, FoscaDiGabiele,AllesandroAlemberti,Nadja Železnik,LeonCizelj-Italy ARCADIA project contribution to the regional cooperation on LFR technology development

111 Siniša Cimeša-Slovenia Transposition challenges of new WENRA requirements into Slovenian regulation

112 Tomaž Skobe-Slovenia Quality Assurance System in Nuclear Training Centre

113 Pavel Gabriel Lazaro-Romania Main European Union Citizens’ Attitude Influencers with Respect to Nuclear Energy

114 Jure Hribar,LukaTavčar,MatjažKoželj-Slovenia Analysis of radiation protection training results since adoption of new regulation

116 Nikola Popov-Macedonia Planning of Energy Demand in Macedonia Using the MAED and MESSAGE Model

Thermal Hydraulics 208 Ali Tiftikci-Turkey Turbulent flow simulations of wire-wrapped fuel pin bundle of sodium cooled fast reactor in lattice-Boltzmann framework

209 Simon Adu,IvanHorvatovic,FrancescoD’Auria,BenjaminB.J.BNyarko, Emi-ReynoldsGeoffrey,OforDarkoEmmanuel-Ghana Analysis of Channel Blockage of MNSR Reactor Using the System Thermal- Hydraulic Code RELAP5/MOD3.3

210 Diana Laura Icleanu-Romania Researches Made in Order to Estimate the Implications of Temperature Variations from the Spent Fuel Bay over the Corrosion Rate of the Spent Fuel Cladding Elements

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24th International Conference Nuclear Energy for New Europe xiii

211 Miha Pogačar,IvoKljenak,MatejTekavčič-Slovenia Simulation of Turbulent Liquid Metal Flow in a Triangular Rod Bundle Sub - Channel

212 Paolo Battistoni,MarcoSumini,SandroManservisi,ChristianGonnier, LionelFerry,DidierTarabelli-Italy Preliminary study of the LORELEI test device with the CATHARE-2 code

214 Davide Rozzia,GiuseppeFasano,MarianoTarantino,AlessandroDelNevo, NicolaForgione,AlessandroAlemberti-Italy Experimental Investigation on Powder Conductivity for the Application to Double Wall Bayonet Tube Bundle Steam Generator

215 Ovidiu-Adrian Berar,AndrejProšek,BorutMavko-Slovenia Steady-State Calculation of Krško NPP TRACE model with Three Dimensional Pressure Vessel

217 Il Woong Park,MariaFernandino,CarlosDorao-Norway Effect of the mass flow rate and the subcooling temperature on pressure drop oscillations in a horizontal pipe

218 SafaMohamedAidaroosSalemAlhashmi,HoJoonYoon,Yacine Addad -UnitedArabEmirates Downcomer boiling phenomena analysis during large break loss of coolant accident in APR1400

219 Jure Oder,IztokTiselj-Slovenia Spectral element direct numerical simulation of heat transfer in turbulent channel sodium flow

220 Jure Jazbinšek-Slovenia Modeling of NEK Containment in computer code APROS

221 Sergii Lutsanych,FabioMoretti,NusretAksan,FrancescoD’Auria, AlessandroPetruzzi-Italy FONESYS and SILENCE Networks: Looking to the Future of T-H Code Development and Experimentation

222 Hüseyin Ayhan,CemalNiyaziSökmen-Turkey Determination of Geometrical and Operating Parameters of PRHR for VVER Reactors: Cooling by Natural Circulation of Atmospheric Air

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xiv 24th International Conference Nuclear Energy for New Europe

Research Reactors 303 Ahmad Lashkari - Iran Loss of flow accident analyses in Tehran Research Reactor

305 K.Mayer,MarioVilla,Helmuth Böck-Austria Long-term system outage survey of the TRIGA reactor Vienna

306 Vid Merljak,AndrejTrkov-Slovenia JSI TRIGA 3D Reactor Model – Transition to Cartesian Geometry and Sample Kinetic Simulations

307 Žiga Štancar,LoicBarbot,LukaSnoj-Slovenia Analysis of the TRIGA Mark II Research Reactor Ex-Core Detector Response

309 Junoš Lukan,LukaSnoj-Slovenia Assessing Field Homogeneity: Application To Gamma Radiation Field Around Irradiated Nuclear Fuel

310 Andreas Ikonomopoulos,MelpomeniVarvayanni,NicolasCatsaros-Greece Instrumentation and Control Implementations in Research Reactors: A Review

Reactor Physics 411 Mario Matijević,DavorGrgić,DubravkoPevec-Croatia Severe accident gamma dose distribution through NPP Krško containment and Auxiliary Building calculated using SCALE6.1/MAVRIC sequence

412 IlkemAydogan,Ayhan Yilmazer-Turkey Use of LiF-TLD100 Detector with B4C filter in Neutron Dosimetry

413 Dušan Ćalić,AndrejTrkov-Slovenia Simulation of fuel cycle for Krško NPP using Monte Carlo code and GNOMER diffusion code

414 Paulina Dučkić,KrešimirTrontl,DubravkoPevec-Croatia Application of Support Vector Regression Method on Neutron Buildup Factors

415 Stefan Cerba,BranislavVrban,JakubLüley,JanJascik,VladimirNečas-Slovakia Investigation of the Allegro MOX Pin Core design by stochastic and deterministic methods

416 Ahmad Lashkari - Iran Reactivity power and temperature coefficients determination of the TRR

417 Zsolt Soti,MagillJoseph,DreherRaymond,PfennigGerda-Germany The New Edition of Karlsruhe Nuclide Chart in Summer 2015

418 Antonio Guglielmelli,FedericoRocchi,MarcoSumini-Italy Scale 6.1.3 evaluation of the heavy reflector effective cross sections of a GEN III PWR system and a Serpent 2.1.23 model comparison

419 Antonio Guglielmelli,FedericoRocchi,MarcoSumini-Italy Scale 6.1.3 effective heavy reflector cross sections sensitivity analysis for a PWR GENIII assembly/reflector system

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24th International Conference Nuclear Energy for New Europe xv

420 Matjaž Božič,MartinChambers,BojanKurinčič-Slovenia BEACON Version 6 vs 7 Software Code Comparison in NPP Krško Operating Cycle 27

421 Antonella Labarile,TeresaBarrachina,RafaelMiró,GumersindoVerdú-Spain Participation in OECD/NEA Oskarshamn-2 (O2) BWR Stability Benchmark for Uncertainty Analysis in Modelling Using Triton and Keno for Transport Calculations and Tsunami and Sampler for Cross Section Error Propagation

422 Antonella Labarile,TeresaBarrachina,RafaelMiró,GumersindoVerdú-Spain TRITON vs POLARIS. Comparison between two modules for LWRs modelling in SCALE6.2

423 Marjan Kromar,BojanKurinčič,UrbanSimončič,RokBizjak-Slovenia Neutron noise analysis in the NPP Krško - Comparison of Cycles 26, 27 and 28

424 Melpomeni Varvayanni,AndreasIkonomopoulos,NicolasCatsaros-Greece An Overview of the Improvements in Fuel Cycle Sustainability for GEN III+ Reactors

425 Consuelo Gómez-Zarzuela Quel,AgustinAbarca,TeresaBarrachina,RafaMiró, GumersindoVerdú-Spain Ringhals-1 BWR stability analysis with TRACE/PARCS

Severe Accidents & Probabilistic Safety Assessment 506 Mitja Uršič,MatjažLeskovar-Slovenia Potential of vapour explosions in sodium

507 Mitja Antončič,ŽivaBricmanRejc,MarkoČepin-Slovenia Probabilistic Safety Assessment of Shutdown and Refueling States

508 Cemil Kocar,CigdemPolatDagli-Turkey Analysis of LOFA in BWR Spent Fuel Storage Pool

509 Aleksander Grah-Netherlands In vessel melt retention (IVMR) for a VVER 1000 reactor – CFD computation

510 Vasilij Centrih,MatjažLeskovar-Slovenia Analysis of ZrO2/WO3 vs. ZrO2/UO2 Fuel-Coolant Interaction in KROTOS Conditions

511 Aurimas Kontautas-Lithuania Fission Product and Aerosol Deposition Analysis in Phebus Containment Under FPT3

512 Mitja Uršič,MatjažLeskovar-Slovenia Analyses of THINA melt-sodium interaction experiments with MC3D

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xvi 24th International Conference Nuclear Energy for New Europe

Radioactive Waste, Environmental issues 605 Thomas Breznik,MarkoGerbec,BorutSmodiš-Slovenia Barriers and Operational Risk Assessment of Incidents and Accidents occurring in the Transport of Radioactive Materials

606 Ljudmila Benedik,MarkoŠtrok,BarbaraSvetek,ZdenkaTrkov-Slovenia Radiochemical techniques for determination of actinides, Po-210, H-3, C-14 and Sr-89/90 in urine samples

608 L. T. Dobrev,B.Slavchev,A.Chalakov-Bulgaria Radioanalytical methods laboratory – past, present and future

609 Radu Secareanu,MinoruTakahashi,RiccardoMereu,IliePrisecaru-Romania Breakup and Solidification Behaviour of Liquid Metal Jet in Water Environment

610 MarkoGordić,DejviKadivnik,BojanKurinčič,Martin Chambers-Slovenia Traveller Implementation and Experiences at NPP Krško

611 Matjaž Koželj,IgorJenčič-Slovenia Radiation Protection Training Needs in Slovenia

Nuclear Fusion 704 Leon Kos,JanezKrek,MarijoTelenta-Slovenia Visualisation of Fusion Related Models Stored in General Grid Description

705 Aleksander Drenik,MartinOberkofler,DanielAlegre,UronKruezi, SebastijanBrezinsek,MarcoWischmeier,CarineGiroud,MiranMozetičand JETContributors-Slovenia Sub-Divertor Neutral Gas Analysis at JET with the ITER-like Wall

706 Iztok Čadež,SabinaMarkelj,AnžeZaložnik-Slovenia Study of some atomic and molecular processes relevant to the tokamak edge plasma modelling

707 Marijo Telenta,LeonKos,RobertAkers-Slovenia Interfacing of CAD models to a Common Fusion Modelling Grid Description

708 Gabrijela Ikovic,LinoŠalamon,TomažGyergyek-Slovenia Simulations of an Ion Energy Analyzer using PIC technique

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24th International Conference Nuclear Energy for New Europe xvii

Materials 807 Samir El Shawish,LeonCizelj-Slovenia Calculation of Intergranular Stress and Strain Distributions in Neutron- Irradiated Stainless Steel Aggregate Model

808 Matej Pleterski,JoškoValentinčič,IzidorSabotin-Slovenia Heat Exchanger Tube Cutting System

809 Francesco Dolci-Netherlands Thermo-mechanical model of a pipe under thermal fatigue

810 Petar Mateljak,EstefaniaArtigao,EleniCheilakou,VassilisKappatos, AlvaroGarcia,MarkoBudimir-Croatia Vortex Robot for Rapid Low Cost Scanning and Improved Non-Destructive Testing of Large Concrete Structures

811 Davide Rozzia,AlessandroDelNevo,LelioLuzzi-Italy Modeling and Assessment of PCI in LWR Fuel

812 Hygreeva Namburi-CzechRepublic Effect of tensile strain on micro-structure of irradiated core internal material

Session 7 16:20 Reactor Physics II Chairpersons: DubravkoPevec,Croatia

MelpomeniVarvayanni,Greece

16:20 406 CihangirÇelik,Mehmet Tombakoglu-USA Application of Monte Carlo Method for Burnup Dependent Full Core Neutronic Analysis of PBMR

16:40 407 Antonios Mylonakis,MelpomeniVarvayanni,NicolasCatsaros-Greece Investigating a Newton-based, matrix-free, Neutronic-Monte Carlo/Thermal Hydraulic coupling scheme

17:00 408 György Hegyi,CsabaMaráczy,GáborHordósy,EmeseTemesvári-Hungary Evaluation of the Full Core VVER-440 Benchmarks Using the KARATE and MCNP Code Systems

17:20 409 ThaliaXenofontos,GregoryDelipei,PanayiotaSavva,Melpomeni Varvayanni, JacquesMailliard,NicolasCatsaros,B.Gaveau-Greece ANET Reaction Rates Validation Based on the VENUS-2 MOX Core Benchmark Analysis

17:40 410 Marjan Kromar,BojanKurinčič-Slovenia Determination of the NPP Krško Spent Fuel Activity

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xviii 24th International Conference Nuclear Energy for New Europe

Wednesday, September 16

Session 8 08:30 Severe Accidents & Probabilistic Safety Assessment Chairpersons: GerardCognet,France

MatjažLeskovar,Slovenia

08:30 501 LuisE.Herranz,Tim Haste,TeemuKärkelä-Spain The Latest Results from Source Term Research: Overview and Outlook

08:50 502 Maciej Skrzypek,EleonoraSkrzypek,LaurentSaas,RomainLeTellier-Poland In Vessel Corium Propagation Sensitivity Study Of Reactor Pressure Vessel Rupture Time With PROCOR Platform

09:10 503 Jose M. Izquierdo,JavierHortal,EnriqueMeléndez,MiguelSánchez-Spain CSN Experience in the Development and Application of a Computer Platform to Verify Consistency of Deterministic and Probabilistic Arguments in Licensing Safety Cases

09:30 504 Tadej Holler,EdKomen,IvoKljenak-Slovenia Simulation of Hydrogen Combustion Experiment in Large-Scale Experimental Facility with ANSYS Fluent CFD Code

09:50 505 Mantas Povilaitis,St.Kelm,EgidijusUrbonavičius-Lithuania Uncertainty and sensitivity analysis of the Generic Containment SB-LOCA accident

Session 9 10:10 Poster Session Posters Session will also be held on Tuesday at 15:40.

All posters are presented on pages xi - xvii.

Session 10 11:10 Thermal Hydraulics II Chairpersons: IvoKljenak,Slovenia

FrancescoD’Auria,Italy

11:10 204 Blaž Mikuž,IztokTiselj-Slovenia Accurate wall-resolved Large Eddy Simulation of a turbulent flow in 5×5 fuel rod bundle

11:30 205 Sergii Lutsanych,FabioMoretti,FrancescoD’Auria-Italy Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel Experiments

11:50 206 Yacine Addad,AliaMohamedAhmedHammadAlghafri-UnitedArabEmirates Numerical Investigation to Examine Dust impacts on the Dry Cask passive Cooling under U.A.E. harsh environmental conditions

12:10 207 VesnaBenčik,Davor Grgić,SinišaŠadek,NikolaČavlina-Croatia NPP Krško DVI LOCA Calculation Using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence

Program

24th International Conference Nuclear Energy for New Europe xix

Thursday, September 17

Invited Lecture

08:30 3 Karl Krieger,JETContributors-Germany Key results and future directions of the JET fusion research programme

Session 11 09:10 Nuclear Fusion Chairpersons: BoštjanKončar,Slovenia

PrimožPelicon,Slovenia

09:10 701 Sabina Markelj,AnžeZaložnik,ThomasSchwarz-Selinger,MitjaKelemen, PrimožVavpetič,PrimožPelicon-Slovenia Deuterium retention studies in self-ion damaged tungsten exposed to neutral atoms

09:30 702 Lino Šalamon,GabrijelaIkovic,JernejKovačič-Slovenia Ball-pen probe diagnostics of a weakly magnetized discharge plasma column

09:50 703 Oriol Costa Garrido,BoštjanKončar,SamoKošmrlj,ChristianBachmann, BotondMeszaros-Slovenia Global Thermal Analysis of Demo Tokamak

Session 12 10:30 Materials Chairpersons: LeonCizelj,Slovenia

NikolaPopov,Macedonia

10:30 801 Igor Simonovski,TuncayYalcinkaya-Netherlands Strain gradient crystal plasticity approach to modelling micro-plastic flow and localisation in polycrystalline materials

10:50 802 IvanVican,ChannaNageswaran,NikosMakris,AlvaroGarcia,AbbasMohimi, StephanMichau,Marko Budimir-Croatia High Temperature Pipe Structural Health Monitoring System utilising Phased Array probes on TOFD configuration

11:10 803 Fosca Di Gabriele,AlessandroGessi,PetraBublikova,HanaJirkova, DagmarBublikova-CzechRepublic Characterisation of coatings evaluated for LFR applications

11:30 804 AlyAhmed,Davide Rozzia,AlessandroDelNevo,ChristopheDemaziere-Sweden On the effect of MOX fuel conductivity in predicting melting in FR fresh fuel by means of TRANSURANUS code

11:50 805 Florian Haurais,EmilieBeuzet,MartinSteinbrück,YunxiaoWu,AntoineAmbard, EricSimoni,MohamedTorkhani-France Porosimetry of ZrO2 Scales Formed During Oxidation of Zr-Based Fuel Claddings in Nuclear Severe Accident Conditions

12:10 806 Mihai Nicolae Anghel,MarianCuruia,AdrianBadea-Romania Detailed modelling of the thermal radiation shields for applications in superconducting magnetic energy storage systems

Contents

xx 24th International Conference Nuclear Energy for New Europe

Disclaimer

The content of abstracts published in the book of abstracts is the responsibility of the authors concerned. The organizer is not responsible for published facts and technical accuracy of the presented data. The organizer would also like to apologize for any possible errors caused by material processing.

Contents

24th International Conference Nuclear Energy for New Europe xxi

Contents1 Sustainable Electricity: Wishful thinking or near-term reality?

StefanHirschberg 12 The analysis of crud light water reactor fuel rods, and its thermal hydraulic consequences

SimonP.Walker 23 Key results and future directions of the JET fusion research programme

KarlKrieger,JETContributors 3101 Nuclear Energy, Challenges for the Future

ChristopheBéhar,PierreLeCoz 4102 The need of development Gas Cooled Reactor Technology in Europe

AnnaPrzybyszewska,KajetanRozycki 5103 Human Resources Infrastructure Requirements for New Nuclear Power Program

NikolaPopov 6104 How people perceive ionizing radiation: comparison in four countries

NadjaŽeleznik 6105 The Role of High Performance Computing in the Nuclear Energy Sector

IgorSimonovski,BojanŽefran,SteveClements,SandiCimerman 7106 Use of the Geiger-Müller counter and the cloud chamber to present properties of radioactivity to youngsters

VesnaSlaparBorišek 8107 Public Opinion about Nuclear Energy – Year 2015 Poll

RadkoIstenič,IgorJenčič 8108 Proposal of a BEPU-FSAR

FrancineMenzel,FrancescoSaverioD'Auria,GaianeSabundjian, AlziraAbrantesMadeira 9

109 Inputs for national research strategies for coordination of social, societal and governance issues in nuclear energy

NadjaŽeleznik,KjellAndersson 9110 ARCADIA project contribution to the regional cooperation on LFR technology development

DanielaDiaconu,MarinConstantin,GeorgiosGlinatsis,GiacomoGrasso, FoscaDiGabiele,AllesandroAlemberti,NadjaŽeleznik,LeonCizelj 10

111 Transposition challenges of new WENRA requirements into Slovenian regulation SinišaCimeša 11

112 Quality Assurance System in Nuclear Training CentreTomažSkobe 11

113 Main European Union Citizens’ Attitude Influencers with Respect to Nuclear Energy

PavelGabrielLazaro 12114 Analysis of radiation protection training results since adoption of new regulation

JureHribar,LukaTavčar,MatjažKoželj 13115 Nuclear Power Infrastructure Assessment and Regional Cooperation in Macedonia

NikolaPopov 13116 Planning of Energy Demand in Macedonia Using the MAED Model

NikolaPopov 14

Contents

xxii 24th International Conference Nuclear Energy for New Europe

201 Application of ISA methodology to a Loss of Normal Feedwater ATWS with TRACE 5.0

Maria-JoseRebollo 15202 Uncertainty Quantification of NEPTUNE_CFD calculation by Optimal Statistical Estimator Method

AndrejProšek,BoštjanKončar,MatjažLeskovar 15204 Accurate wall-resolved Large Eddy Simulation of a turbulent flow in 5×5 fuel rod bundle

BlažMikuž,IztokTiselj 16205 Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel Experiments

SergiiLutsanych,FabioMoretti,FrancescoSaverioD'Auria 17206 Numerical Investigation to Examine Dust impacts on the Dry Cask passive Cooling under U.A.E. harsh environmental conditions

YacineAddad,AliaMohamedAhmedHammadAlghafri 17207 NPP Krško DVI LOCA Calculation Using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence

VesnaBenčik,DavorGrgić,SinišaŠadek,NikolaČavlina 18208 Turbulent flow simulations of wire-wrapped fuel pin bundle of sodium cooled fast reactor in lattice-Boltzmann framework

AliTiftikci,CemilKocar 19209 Analysis of Channel Blockage of MNSR Reactor Using the System Thermal-Hydraulic Code RELAP5/MOD3.3

SimonAdu,IvanHorvatovic,FrancescoSaverioD'Auria,BenjaminB.J.BNyarko, Emi-ReynoldsGeoffrey,OforDarkoEmmanuel 19

210 Researches Made in Order to Estimate the Implications of Temperature Variations from the Spent Fuel Bay over the Corrosion Rate of the Spent Fuel Cladding Elements

DianaLauraIcleanu 20211 Simulation of Turbulent Liquid Metal Flow in a Triangular Rod Bundle Sub - Channel

MihaPogačar,IvoKljenak,MatejTekavčič 20212 Preliminary study of the LORELEI test device with the CATHARE-2 code

PaoloBattistoni,MarcoSumini,SandroManservisi,ChristianGonnier,LionelFerry, DidierTarabelli 21

213 Influence of Liquid Inlet Modeling on Simulated Wave Characteristics in Vertical Gas-Liquid Churn Flow

MatejTekavčič,BoštjanKončar,IvoKljenak 21214 Experimental Investigation on Powder Conductivity for the Application to Double Wall Bayonet Tube Bundle Steam Generator

DavideRozzia,GiuseppeFasano,MarianoTarantino,AlessandroDelNevo, NicolaForgione,AlessandroAlemberti 22

215 Steady-State Calculation of Krško NPP TRACE model with Three Dimensional Pressure Vessel

Ovidiu-AdrianBerar,AndrejProšek,BorutMavko 23217 Effect of the mass flow rate and the subcooling temperature on pressure drop oscillations in a horizontal pipe

IlWoongPark,MariaFernandino,CarlosDorao 23218 Downcomer boiling phenomena analysis during large break loss of coolant accident in APR1400

SafaMohamedAidaroosSalemAlhashmi,HoJoonYoon,YacineAddad 24

Contents

24th International Conference Nuclear Energy for New Europe xxiii

219 Spectral element direct numerical simulation of heat transfer in turbulent channel sodium flow

JureOder,IztokTiselj 25220 Modeling of NEK Containment in computer code APROS

JureJazbinšek 25221 FONESYS and SILENCE Networks: Looking to the Future of T-H Code Development and Experimentation

SergiiLutsanych,FabioMoretti,NusretAksan,FrancescoSaverioD'Auria, AlessandroPetruzzi 26

222 Determination of Geometrical and Operating Parameters of PRHR for VVER Reactors: Cooling by Natural Circulation of Atmospheric Air

HüseyinAyhan,CemalNiyaziSökmen 27301 Human Resource Development and Nuclear Education Through Research Reactors: Successful Approach to Build Up the Future Generation of Nuclear Professionals

HelmuthBöck,MarioVilla,AndreaBorioDiTigliole,JudyVyshniauskas, LubomirSklenka,LukaSnoj,AttilaTormási 28

302 Measurements with Multiple In-core Fission Chambers at the JSI TRIGA Mark II Reactor

TanjaKaiba,GašperŽerovnik,LukaSnoj 29303 Loss of flow accident analyses in Tehran Research Reactor

AhmadLashkari 29304 Analysis of coolant temperature distribution for the validation of TRIGA Mark II CFD computational model

RomainHenry,MarkoMatkovič 30305 Long-term system outage survey of the TRIGA reactor Vienna

K.Mayer,MarioVilla,HelmuthBöck 30306 JSI TRIGA 3D Reactor Model – Transition to Cartesian Geometry and Sample Kinetic Simulations

VidMerljak,AndrejTrkov 31307 Analysis of the TRIGA Mark II Research Reactor Ex-Core Detector Response

ŽigaŠtancar,LoicBarbot,LukaSnoj 31309 Assessing Field Homogeneity: Application To Gamma Radiation Field Around Irradiated Nuclear Fuel

JunošLukan,LukaSnoj 32310 Instrumentation and Control Implementations in Research Reactors: A Review

AndreasIkonomopoulos,MelpomeniVarvayanni,NicolasCatsaros 33401 Simulation of the NPP Krško Core at Hot Full Power with CASL Core Simulator - VERA-CS

FaustoFranceschini,MarjanKromar,AndrewT.Godfrey 34402 Experimental study of the physical properties of ADS systems – measurement of high energy neutron fields by using the 89Y threshold detectors

MarcinBielewicz,ElżbietaStrugalska-Gola,StanisławKilim,MarcinSzuta 34403 Sensitivity Analysis of Gas-cooled Fast Reactor

JakubLüley,StefanCerba,BranislavVrban,JánHaščík,VladimirNečas 35404 Preliminary Analysis of The FLUOLE-2 Experiment

StephaneBourganel,JacquesDi-Salvo,ThiollayNicolas,SoldevilaMichel 35405 ALLEGRO uncertainty and similarity evaluation

BranislavVrban,StefanCerba,JakubLüley,JánHaščík,VladimirNečas 36

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xxiv 24th International Conference Nuclear Energy for New Europe

406 Application of Monte Carlo Method for Burnup Dependent Full Core Neutronic Analysis of PBMR

CihangirÇelik,MehmetTombakoglu 37407 Investigating a Newton-based, matrix-free, Neutronic-Monte Carlo/Thermal Hydraulic coupling scheme

AntoniosMylonakis,MelpomeniVarvayanni,NicolasCatsaros 37408 Evaluation of the Full Core VVER-440 Benchmarks Using the KARATE and MCNP Code Systems

GyörgyHegyi,CsabaMaráczy,GáborHordósy,EmeseTemesvári 38409 ANET Reaction Rates Validation Based on the VENUS-2 MOX Core Benchmark Analysis

ThaliaXenofontos,GregoryKyriakosDelipei,PanayiotaSavva, MelpomeniVarvayanni,JacquesMailliard,NicolasCatsaros,B.Gaveau 39

410 Determination of the NPP Krško Spent Fuel Activity MarjanKromar,BojanKurinčič 39

411 Severe accident gamma dose distribution through NPP Krško containment and Auxiliary Building calculated using SCALE6.1/MAVRIC sequence

MarioMatijević,DavorGrgić,DubravkoPevec 40412 Use of LiF-TLD100 Detector with B4C filter in Neutron Dosimetry

IlkemAydogan,AyhanYilmazer 40413 Simulation of fuel cycle for Krško NPP using Monte Carlo code and GNOMER diffusion code

DušanĆalić,AndrejTrkov 41414 Application of Support Vector Regression Method on Neutron Buildup Factors

PaulinaDučkić,KrešimirTrontl,DubravkoPevec 41415 Investigation of the Allegro MOX Pin Core design by stochastic and deterministic methods

StefanCerba,BranislavVrban,JakubLüley,JanJascik,VladimirNečas 42416 Reactivity power and temperature coefficients determination of the TRR

AhmadLashkari 43417 The New Edition of Karlsruhe Nuclide Chart in Summer 2015

ZsoltSoti,MagillJoseph,DreherRaymond,PfennigGerda 43418 Scale 6.1.3 evaluation of the heavy reflector effective cross sections of a GEN III PWR system and a Serpent 2.1.23 model comparison

AntonioGuglielmelli,FedericoRocchi,MarcoSumini 44419 Scale 6.1.3 effective heavy reflector cross sections sensitivity analysis for a PWR GENIII assembly/reflector system

AntonioGuglielmelli,FedericoRocchi,MarcoSumini 45420 BEACON Version 6 vs 7 Software Code Comparison in NPP Krško Operating Cycle 27

MatjažBožič,MartinChambers,BojanKurinčič 45421 Participation in OECD/NEA Oskarshamn-2 (O2) BWR Stability Benchmark for Uncertainty Analysis in Modelling Using Triton and Keno for Transport Calculations and Tsunami and Sampler for Cross Section Error Propagation

AntonellaLabarile,TeresaBarrachina,RafaelMiró,GumersindoVerdú 46422 TRITON vs POLARIS. Comparison between two modules for LWRs modelling in SCALE6.2

AntonellaLabarile,TeresaBarrachina,RafaelMiró,GumersindoVerdú 47423 Neutron noise analysis in the NPP Krško - Comparison of Cycles 26, 27 and 28

MarjanKromar,BojanKurinčič,UrbanSimončič,RokBizjak 47

Contents

24th International Conference Nuclear Energy for New Europe xxv

424 An Overview of the Improvements in Fuel Cycle Sustainability for GEN III+ Reactors

MelpomeniVarvayanni,AndreasIkonomopoulos,NicolasCatsaros 48425 Ringhals-1 BWR stability analysis with TRACE/PARCS

ConsueloGómez-ZarzuelaQuel,AgustinAbarca,TeresaBarrachina,RafaMiró,GumersindoVerdú 49

501 The Latest Results from Source Term Research: Overview and OutlookLuisE.Herranz,TimHaste,TeemuKärkelä3 50

502 In Vessel Corium Propagation Sensitivity Study Of Reactor Pressure Vessel Rupture Time With PROCOR Platform

MaciejSkrzypek,EleonoraKlaraSkrzypek,LaurentSaas,RomainLeTellier 51503 CSN Experience in the Development and Application of a Computer Platform to Verify Consistency of Deterministic and Probabilistic Arguments in Licensing Safety Cases

JoseM.Izquierdo,JavierHortal,EnriqueMeléndez,MiguelSánchez 51504 Simulation of Hydrogen Combustion Experiment in Large-Scale Experimental Facility with ANSYS Fluent CFD Code

TadejHoller,EdKomen,IvoKljenak 52505 Uncertainty and sensitivity analysis of the Generic Containment SB-LOCA accident

MantasPovilaitis,St.Kelm,EgidijusUrbonavičius 53506 Potential of vapour explosions in sodium

MitjaUršič,MatjažLeskovar 53507 Probabilistic Safety Assessment of Shutdown and Refueling States

MitjaAntončič,ŽivaBricmanRejc,MarkoČepin 54508 Analysis of LOFA in BWR Spent Fuel Storage Pool

CemilKocar,CigdemPolatDagli 54509 In vessel melt retention (IVMR) for a VVER 1000 reactor – CFD computation

AleksanderGrah 55510 Analysis of ZrO2/WO3 vs. ZrO2/UO2 Fuel-Coolant Interaction in KROTOS Conditions

VasilijCentrih,MatjažLeskovar 55511 Fission Product and Aerosol Deposition Analysis in Phebus Containment Under FPT3

AurimasKontautas 56512 Analyses of THINA melt-sodium interaction experiments with MC3D

MitjaUršič,MatjažLeskovar 56601 Preparation of the national program for the spent fuel and radioactive waste management taking into account possibility of European Repository Development Organisation development

TomažŽagar,LeonKegel 57602 On-line relative air dispersion concentrations one week forecast for Krško NPP prepared for routine and emergency use

PrimožMlakar,BoštjanGrašič,MarijaZlataBožnar,BorutBreznik2 57603 Transfer of Th-230 from soil contaminated with U-mill tailing to radish, savoy and rocket

PetraPlaninšek,BorutSmodiš,LjudmilaBenedik 58605 Barriers and Operational Risk Assessment of Incidents and Accidents occurring in the Transport of Radioactive Materials

ThomasBreznik,MarkoGerbec,BorutSmodiš 59

Contents

xxvi 24th International Conference Nuclear Energy for New Europe

606 Radiochemical techniques for determination of actinides, Po-210, H-3, C-14 and Sr-89/90 in urine samples

LjudmilaBenedik,MarkoŠtrok,BarbaraSvetek,ZdenkaTrkov 60608 Radioanalytical methods laboratory – past, present and future

L.T.Dobrev,B.Slavchev,A.Chalakov 61609 Breakup and Solidification Behaviour of Liquid Metal Jet in Water Environment

RaduSecareanu,MinoruTakahashi,RiccardoMereu,IliePrisecaru 62610 Traveller Implementation and Experiences at NPP Krško

MarkoGordić,DejviKadivnik,BojanKurinčič,MartinChambers 62611 Radiation Protection Training Needs in Slovenia

MatjažKoželj,IgorJenčič 63701 Deuterium retention studies in self-ion damaged tungsten exposed to neutral atoms

SabinaMarkelj,AnžeZaložnik,ThomasSchwarz-Selinger,MitjaKelemen, PrimožVavpetič,PrimožPelicon 64

702 Ball-pen probe diagnostics of a weakly magnetized discharge plasma column LinoŠalamon,GabrijelaIkovic,JernejKovačič 65

703 Global Thermal Analysis of Demo TokamakOriolCostaGarrido,BoštjanKončar,SamoKošmrlj,ChristianBachmann, BotondMeszaros 65

704 Visualisation of Fusion Related Models Stored in General Grid DescriptionLeonKos,JanezKrek,MarijoTelenta 66

705 Sub-Divertor Neutral Gas Analysis at JET with the ITER-like WallAleksanderDrenik,MartinOberkofler,DanielAlegre,UronKruezi, SebastijanBrezinsek,MarcoWischmeier,CarineGiroud,MiranMozetičand JETContributors 67

706 Study of some atomic and molecular processes relevant to the tokamak edge plasma modelling

IztokČadež,SabinaMarkelj,AnžeZaložnik 68707 Interfacing of CAD models to a Common Fusion Modelling Grid Description

MarijoTelenta,LeonKos,RobertAkers 68708 Simulations of an Ion Energy Analyzer using PIC technique

GabrijelaIkovic,LinoŠalamon,TomažGyergyek 69801 Strain gradient crystal plasticity approach to modelling micro-plastic flow and localisation in polycrystalline materials

IgorSimonovski,TuncayYalcinkaya 70802 High Temperature Pipe Structural Health Monitoring System utilising Phased Array probes on TOFD configuration

IvanVican,ChannaNageswaran,NikosMakris,AlvaroGarcia,AbbasMohimi, StephanMichau,MarkoBudimir 70

803 Characterisation of coatings evaluated for LFR applicationsFoscaDiGabriele,AlessandroGessi,PetraBublikova,HanaJirkova, DagmarBublikova 71

804 On the effect of MOX fuel conductivity in predicting melting in FR fresh fuel by means of TRANSURANUS code

AlyAhmed,DavideRozzia,AlessandroDelNevo,ChristopheDemaziere 72

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24th International Conference Nuclear Energy for New Europe xxvii

805 Porosimetry of ZrO2 Scales Formed During Oxidation of Zr-Based Fuel Claddings in Nuclear Severe Accident Conditions

FlorianHaurais,EmilieBeuzet,MartinSteinbrück,YunxiaoWu,AntoineAmbard, EricSimoni,MohamedTorkhani 73

806 Detailed modelling of the thermal radiation shields for applications in uperconducting magnetic energy storage systems

MihaiNicolaeAnghel,MarianCuruia,AdrianBadea 74807 Calculation of Intergranular Stress and Strain Distributions in Neutron-Irradiated Stainless Steel Aggregate Model

SamirElShawish,LeonCizelj 74808 Heat Exchanger Tube Cutting System

MatejPleterski,JoškoValentinčič,IzidorSabotin 75809 Thermo-mechanical model of a pipe under thermal fatigue

FrancescoDolci 75810 Vortex Robot for Rapid Low Cost Scanning and Improved Non-Destructive Testing of Large Concrete Structures

PetarMateljak,EstefaniaArtigao,EleniCheilakou,VassilisKappatos,AlvaroGarcia,MarkoBudimir 76

811 Modeling and Assessment of PCI in LWR FuelDavideRozzia,AlessandroDelNevo,LelioLuzzi 77

812 Effect of tensile strain on micro-structure of irradiated core internal materialHygreevaKiranNamburi 77

Contents

xxviii 24th International Conference Nuclear Energy for New Europe

Invited Lectures

24th International Conference Nuclear Energy for New Europe 1

Invited Lectures1

Sustainable Electricity: Wishful thinking or near-term reality?Stefan HirschbergPaulScherrerInstitut-PSILaboratoryforEnergySystemsAnalysis,EnergyDepartments,CH-5232VilligenPSI,[email protected]

Striving for sustainability enjoys a wide support. It is easy to agree that energy technologies that are climate friendly, have small impacts on human health and ecosystems, are characterized by low accident risks, are resource-saving, exhibit high reliability of supply, are economically affordable and have broad social acceptance come close to satisfying the basic sustainability requirements. Are such technologies available today? If not, will they be available in short-, middle- or long-term? Are there transparent, scientific approaches that allow us to measure the sustainability of the various options?The presentation highlights insights from comparative studies of fossil, nuclear and renewable electricity supply systems using a framework for systematic comparative evaluation of energy systems. This aims to improve transparency and the systematic use of the objective knowledge base and allows crucial aspects of the various energy sources to be addressed in relation to economic, environmental and social dimensions. In particular results from recent comprehensive assessments are provided, illustrating strengths and weaknesses of the various current and future electricity supply technology options.Major progress has been made by the Paul Scherrer Institute (PSI) in cooperation with research partners in the development and implementation of a framework for sustainability assessment of electricity supply technologies with the associated fuel cycles. Consistent, measurable, technology-specific indicators for environmental, economic and social performance must be considered, for example pollutant emissions, generation costs, and the consequences of possible accidents. The representative quantitative indicators are derived, using a variety of methods including Life Cycle Assessment, Impact Pathway Approach and Probabilistic Safety Assessment. The methodological framework is supported by large databases. Using Multi-Criteria Decision Analysis (MCDA), the measured indicators can be combined with subjective preferences of the various stakeholders. Recent applications cover a variety of cases such as Switzerland, selected European Union countries and China. The work demonstrates that the indicator-based assessment is highly suitable for guiding in a structured manner the debate on the future energy supply and for supporting informed decisions. Sensitivity patterns with regard to such preferences are shown.

Keywords: sustainability, quantification, indicators, stakeholders, preferences

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Invited Lectures

2 24th International Conference Nuclear Energy for New Europe

2

The analysis of crud light water reactor fuel rods, and its thermal hydraulic consequences

Simon P. WalkerDepartmentofMechanicalEngineering,ImperialCollege,ExhibitionRoad,London,SW72BX,[email protected]

The water coolant in PWR and BWR systems contains both dissolved ionic species and suspended particulates. Some of these are deliberately added, whilst others are the result of corrosion of the iron and nickel-based alloys used to construct the primary circuit. Some of these materials can leave solution and form a solid deposit on the surface of the fuel cladding, particularly where boiling occurs. Such deposits, termed “crud”, are usually highly porous, consist of a variety of metal oxides and other components and can be anything from 1 to 100μm thick. Fuel crud has a variety of undesirable effects on the performance of the plant. These include: crud-induced power shifts (CIPS), crud induced localised corrosion (CILC), increased ex-core operator doses and potential long term fuel storage problems. There are therefore both cost and safety benefits to understanding and controlling this phenomenon. Current nuclear plants are increasing their power output (e.g. the Ringhals PWR plant in Sweden carried out an 18% uprate in 2014) and new designs are based on high power density cores, so it is likely the problem of fuel crud will increase in the future.This talk will report:(i) The ‘macroscopic’ analysis of the thermal hydraulic consequences of crud, for flow diversion, clad temperature, and DNBR, and(ii) The development of a microscopic understanding of the processes at work with the crud, attempting to gain insight into the heat transfer and chemistry behaviour.It will finish with a short description of a planned extension of this detailed modelling to address the ‘chemistry’ issues in more detail.

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Invited Lectures


3

Key results and future directions of the JET fusion research programme

Karl Krieger1, JET Contributors*2

1 Max-Planck-InstitutfürPlasmaphysik(IPP),Boltzmannstr.2,D-85748Garching,Germany2 EUROfusionConsortium,JET,CulhamScienceCentre,OX143DB,Abingdon,[email protected]

The JET research programme is a key element of the EUROfusion programme in support of the ITER experimental fusion reactor presently under construction in France. For that purpose the JET plasma facing in vessel components were changed to an ITER-like wall configuration (ILW) made of the same materials (Be, W) as will be used in ITER. In the first ILW exploitation phase from 2011-14 the compatibility of ITER plasma operation scenarios with the new all metal wall was successfully demonstrated. Moreover, the predicted reduction of fuel retention for the ILW by an order of magnitude, compared to the previously used carbon wall, could be confirmed. The upcoming JET programme for 2015-2016 is focused on achieving maximum plasma performance in deuterium plasmas for the two main ITER regimes of plasma operation. Based on this optimisation process, the final phase of JET experimentation, presently scheduled for 2017, will demonstrate plasma operation in these regimes with deuterium-tritium fusion fuel mix at the (scaled) performance level required for ITER. The corresponding generation of several MWs of fusion power will allow studying the physics processes of fusion alpha particle heating and their influence on plasma properties. In addition the use of tritium will allow to further explore the known dependencies of plasma processes and properties on the fuel isotope mass. Apart from the physics research programme the JET D-T experiments will also provide an opportunity to test a variety of technologies that will be required in ITER and a fusion reactor, such as tritium handling, fuel cycle and removal, as well as handling of neutron radiation and activation effects and corresponding radiation safety measures.“This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.”

*See App. of F. Romanelli et al., Proc. of the 25th IAEA Fusion Energy Conf. 2014, Saint Petersburg, Russia

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Nuclear Energy and Society


Nuclear Energy and Society101

Nuclear Energy, Challenges for the FutureChristophe Béhar1, Pierre Le Coz2

1 CEA-France,Directiondel'EnergieNucléaire,Bât.121,91191GifsurYvetteCedex,France2 CommissariataL’EnergieAtomiqueDirectiondesreacteursnucleaires,CEA/CADARACHE-BT211, [email protected]

Including nuclear power in a sustainable development plan, presupposes a long-term collective commitment to the safe management of a competitive nuclear energy. It is the meaning of major programmes conducted by CEA’s Nuclear Energy Division in order to increase the current fleet competitiveness, with industrial implications in terms of reactor service life, performance, availability and safety, and optimize or adapt front-end and back-end facilities of the nuclear fuel cycle to meet current and future industrial challenges. It presupposes also the development of solutions for the future, making the best of the resources, and ensuring that all operations meet the most stringent safety criteria challenged for fast neutron reactors of Generation IV.In accordance with the French Act of June 28, 2006, CEA’s Nuclear Energy Division is responsible for designing the technological reactor ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), based on a strong feedback experience and combining thus essential proven features. ASTRID is a 1500 thMW self-sustainable pool-type SFR with its energy conversion system. In addition of its prototype character, ASTRID should authorize some experimental capabilities such as offering irradiation services for demonstrating fast reactors flexibility to breed or burn Plutonium, and transmute minor actinides, advanced fuels and materials. The CEA acts as the industrial architect of the project. ASTRID engineering is conducted within a broad cooperative framework, gathering already 13 industrial partners. It was lately joined by a pool of Japanese partners covering both ASTRID design and R&D in support to the project. At the same time the project gathers several European and international research laboratories to conduct R&D in support of innovation requirements of the project, and takes advantage of the European network ARDECO set up by the CEA. France is determined to succeed and, with its international and European partners, to be one of the first countries with a comprehensive 4th-generation reactor design package.In 2014, the French Safety Advisory Committee for nuclear reactors considered that among the 6 nuclear energy systems selected by the Generation IV International Forum (GIF), the SFR system is the only one with a sufficient level of maturity for the realization of a prototype in the first half of the 21st century. Studies for the second step of the conceptual design phase are ongoing and a final report is expected by the end of 2015 together with the Safety Options File. This foreshadows the forthcoming 2016-2019 phase for basic design and preparation of regulatory documentation, before the detailed design and regulatory process for the construction license.

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102

The need of development Gas Cooled Reactor Technology in Europe

Anna Przybyszewska, Kajetan RozyckiNationalCentreforNuclearResearch,ul.AndrzejaSołtana7,Otwock-Świerk,[email protected]

Use of gas cooling instead of light water is envisaged in many prospective designs of nuclear plants. Nowadays we can observe nowadays several initiatives concerning Gas Cooled Reactors (GCRs) and in particular High Temperature Reactors (HTRs) around the World. The main reasons for this interest are related to advantages of gas cooled reactor technology over light water technology, such as:• Improved efficiency (up to about 45%) due to higher outlet temperatures of cooling medium.• Possible use for supply of process heat to energy intensive industries.• Excellent passive safety, if reactor is properly designed.The National Centre for Nuclear Resarch (NCBJ) in Poland is involved in development of an European platform of knowledge and collaboration - The Sustainable Nuclear Energy Technology Platform (SNETP). Under two of its three pillars projects linked to gas cooled reactor technology are conducted. One of the main advantages of HTRs is their suitability to work in cogeneration mode. Cogeneration could extend the low carbon contribution from nuclear fission to the energy system by directly providing heat for different applications like district heating, sea water desalination, process heat for many industrial applications as well as bulk hydrogen production, synthetic transport fuel production or even carbon capture and utilization (CCU). Potential for deployment of HTR technology to fulfil industrial energy needs created the Nuclear Cogeneration Industrial Initiative (NC2I). NCBJ is a leader of EU FP7 supported project “NC2I-R” which studies feasibility of using nuclear reactors to produce electricity and process heat.Similar activities are conducted at national level. For example, HTR-PL is the project supported at ministerial level with the main objective to increase the research and technical potential and help the development of nuclear energy in Poland. One of the major challenges against broader implementation of nuclear energy is the issue of nuclear waste (spent fuel). Fast spectrum reactors with closed fuel cycles will allow reduction in high-level nuclear waste radiotoxicity and volume. The use of fast reactors with a closed fuel cycle approach, as a matter of fact will also allow more sustainable implementation of nuclear energy. Several coolants may be used in fast reactor. One of them is Helium. ALLEGRO is the second stage of French-led development – also included in an EU FP7 Euratom projects ALLIANCE and ESNII+. A small experimental prototype is envisaged as a first step to prove the technology of fast reactor with gas cooling.The NCBJ is firmly involved in these activities, being therefore visible player in European map of Gas cooled technology.

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103

Human Resources Infrastructure Requirements for New Nuclear Power Program

Nikola PopovELEMMacedonianPowerPlants,11.Octomber9,1000Skopje,[email protected]

Over the past decades nuclear energy has been proven as reliable and economical energy supply that is capable of meeting demanding energy market requirements. Many countries around the world consider entering into new nuclear energy programs and building new power reactors for satisfying their increasing electrical energy needs. Preparations for making a decision to enter into a new nuclear energy program requires a significant amount of human resources, careful planning, preparation and investments.Macedonia is a country with no nuclear reactors (research and power) and the nuclear application is only in medicine, agriculture and food industry, observation for customs needs, radiation measuring in different sectors. On the other side the energy needs has increasing trends in the last ten years, which reflected the electricity import of near 20-30 % (around 3000 GWh). Nuclear power is one of the options for energy expansion planning in the next 50 years. One of the crucial problems in nuclear energy is human resources needs and educational infrastructure in this field. No matter what will be the future energy scenario in Macedonia, the nuclear educational program is the first step to have HR in the field of nuclear energy.The paper will present the proposed direction for having HR in nuclear energy program. Taking into account the existing national education program related to nuclear subjects and nuclear energy, and having the recommended international nuclear educational programs under IAEA, EHRO and national ones, the analyses are made to make the proposed programs for establishing HR infrastructure in nuclear energy field in Macedonia. The other direction will be the establishing the national body and agency related to nuclear energy program in Macedonia.

104

How people perceive ionizing radiation: comparison in four countries

Nadja ŽeleznikRegionalnicenterzaokoljezasrednjoinvzhodnoEvropo,Slovenskacesta5,1000Ljubljana,[email protected]

Investigation of mental models which lay people have regarding the ionizing radiation in several EAGLE partners’ countries has been performed in the frame of EAGLE project. Analyses of mental models in the general public regarding the effects of ionizing radiation aims to examine what are gaps, differences, misunderstandings and misconceptions between professionals in the nuclear area and the public. The main results from the lay people mental models research investigated in four countries (France, Poland, Romania and Slovenia) are:• The attitudes towards the IR, radioactivity and nuclear technology in general slightly depend on age and

gender but mostly dependent of the level and the area of education. Generally, the knowledge about IR is rather low.

• The structure of matter, particularly the structure of atomic nucleus, is rather unclear; therefore the reasons for the decay of a nucleus are very badly known.

• There are many misunderstandings concerning the sources of IR. Often as sources of IR are recognized domestic devices such as microwave oven, cellular phone or TV receiver.

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• The respondents confuse the low, intermediate and high level radioactive waste. Therefore the construction of low and intermediate level radioactive waste repository is not acceptable in the close proximity of the majority’s homes.

• The methods used in nuclear medicine are acceptable, due to trust to the doctors and believes of people that this is another kind of radiation.

• Nuclear power is somehow accepted but not with any great astonishment. • Main media source regarding accidents in nuclear installations are TV, recently also internet. Mass of

internet pages proved different information; sometime obscured or misguiding. Independent sources are appreciated, due to low trust in governmental sources of information.

It has to be emphasized also that the knowledge which was investigated with mental model approach is only one of the dimensions of the communications with public. Many researches also shown that the most important factors are not the one linked with how much people know about ionizing radiation but those linked with perception of risks due to different activity or technology, trust, involvement of the people in the process and opportunities for participation in decision making. This should be constantly take into consideration and also applied in the communication strategies from different sources providing information to the lay population. The research and results will be presented.

105

The Role of High Performance Computing in the Nuclear Energy Sector

Igor Simonovski1, Bojan Žefran2, Steve Clements1, Sandi Cimerman2

1 EuropeanCommission,DGJRC,InstituteforEnergyandTransport,P.O.Box2,NL-1755ZGPetten, Netherlands2 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

High Performance Computing (HPC) plays an ever increasing role in the today’s world. It enables one to model the behaviour of systems, components and structures under various load scenarios, perform complex simulations and test hypotheses. The nuclear energy sector exploits HPC in practically all domains: from research and education to improving nuclear power plant designs, safe operation of nuclear power plants, plant life management for safe long term operation, reducing probability of nuclear accidents and their mitigation. Furthermore, HPC is widely used by academia, research institutes, vendors and authorities.The current paper presents the basic reasons for ever increasing HPC use in the energy field. Evolution of the HPC resources both at the Joint Research Centre of the European Commission and Jožef Stefan Institute, Slovenia, is given. Furthermore, the complexity of successfully setting up such compute resources and providing support to the end-users of the computing facilities is highlighted. The paper concludes with a number of examples of applications in nuclear energy sector and wider.

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106

Use of the Geiger-Müller counter and the cloud chamber to present properties of radioactivity to youngsters

Vesna Slapar BorišekJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

At the Information Centre that was established within the Nuclear Training Centre at the Jožef Stefan Institute visitors are informed about nuclear technology, the basic properties of radioactivity and protection against ionising radiation.The results of the polling of youngsters, conducted over the last 22 years, showed poor understanding of radioactivity. These results are one of the reasons that demonstrations about radioactivity are performed to familiarize visitors with this phenomenon.Demonstrations of the basic properties of the radioactivity are part of each visit, which follow the lecture about nuclear technology. During these demonstrations the Geiger-Müller counter and the cloud chamber are used to show properties of different types of radioactive radiation, using training sources. As a part of demonstrations additional experiments are shown to present the natural radioactivity and the comparison of different types of electromagnetic radiation.

107

Public Opinion about Nuclear Energy – Year 2015 PollRadko Istenič, Igor JenčičJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Public information is an ongoing activity of the Information Centre that was established within the Nuclear Training Centre at the Jožef Stefan Institute more than 20 years ago to. It informs the visitors about nuclear power and nuclear technology in general and about Krško Nuclear Power Plant. The primary target group of information activity are schoolchildren from the 8th and 9th grade of elementary school with their teachers (in total some 8000 per year). The visit consists of a live lecture about nuclear technology followed by the demonstration of radioactivity and a guided tour of a permanent exhibition. Since 1993 we monitor the opinion trends by polling about 1000 youngsters every year. They are polled before they listen to the lecture or visit the exhibition in order to obtain their opinion based on the knowledge from everyday life. In the paper we will present, summarize and comment the trends over the last 22 years.

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108

Proposal of a BEPU-FSAR Francine Menzel1, Francesco Saverio D'Auria2, Gaiane Sabundjian3, Alzira Abrantes Madeira4

1 InstitutodePesquisasEnergéticaseNucleares,SaoPaulo,Brazil2 UniversityofPisa,SanPieroaGradoNuclearResearchGroup(GRNSPG),ViaLivornese1291, 56122Pisa,Italy3 InstitutodePesquisasEnergeticaseNucleares,Av.Prof.LineuPrestes,2242,05508-000SaoPaulo, Brazil4 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,[email protected]

The accident analysis performance consists of a fundamental part of the licensing of the Nuclear Power Plants (NPP). There are conservative and best estimated methods to perform this analysis. Although Best Estimated Plus Uncertainty (BEPU) is used for qualified computational tools and methods of the accident analysis it can be used in other parts of the Final Safety Analysis Report (FSAR), which require Analytical Techniques. Among others the need for uncertainty analysis and the harmonization of the approaches to use the computer codes for these analyses are important things consisting of background to perform a BEPU-FSAR. The objective of this paper is to present the BEPU-FSAR concept and discuss how and why to perform it.

109

Inputs for national research strategies for coordination of social, societal and governance issues in nuclear energy

Nadja Železnik1, Kjell Andersson2

1 RegionalnicenterzaokoljezasrednjoinvzhodnoEvropo,Slovenskacesta5,1000Ljubljana,Slovenia2 KaritaResearchAB,Box6048,SE-18706TÄBY,[email protected]

The development of energy policies, programmes and projects takes place in a social and societal context and these aspects should therefore be an integrated part of national research and development programmes. However, research strategies regarding social, societal and governance aspects of nuclear energy are very rarely addressed even in more developed countries. But they assure coordinated approach with optimisation of related expenses, emphasise the importance of social research on nuclear as an important factor of national decision-making processes concerning the future of nuclear energy and therefore improves the acceptability of associated projects.All the central and eastern European (CEE) countries are currently facing challenges to take certain decisions in the nuclear matters: it might be continuation of the existing nuclear energy sector, building new units or shutting down operating nuclear power plants, or even taking a leading role in the development of new reactor generations. Whichever direction the policy decisions will be taken, a reflection on the social, societal and governance issues should be taken.Within EC PLATENSO project strategies for eight CEE will be prepared and will include:• analysis of the national situation regarding nuclear energy from a societal point of view, • the main objectives and goals with regard to nuclear development, and • measures for fulfilling these including available funds and human resources, time dependencies and

necessary support.

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The strategies will be based on the developed possible scenarios where three different evolutions of events have been looked at and important social topics for research have been identified: continuation of the current situation with reactors in operation and reactors planned to be built in respective countries, phasing out of nuclear power and introduction of new NPPs with Generation 4 reactors. The presentation will give overview of current findings and future work.

110

ARCADIA project contribution to the regional cooperation on LFR technology development

Daniela Diaconu1, Marin Constantin2, Georgios Glinatsis3, Giacomo Grasso4, Fosca Di Gabiele5, Allesandro Alemberti6, Nadja Železnik7, Leon Cizelj8

1 InstituteforNuclearResearchPitesti,CampuluiStreet1,Mioveni,115400,Romania2 INRA-CentredeRecherchesdeNancyLaboratoirePollutionAtmosphérique,F-54280Chapenoux, France3 ENEA,TechnicalUnitforReactorSafetyandFuelCycleMethods,ViaMartiridiMonteSole4, 40129Bologna,Italy4 ENEA,ViaMartiridiMonteSole4,40129Bologna,Italy5 ResearchcentreRez,Hlavni130,25068Husinec-Řež,CzechRepublic6 ANSALDOEnergiaS.p.A.DivisioneNucleare,C.soPerrone25,16161GENOVA,Italy7 RegionalnicenterzaokoljezasrednjoinvzhodnoEvropo,Slovenskacesta5,1000Ljubljana,Slovenia8 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

The Union Energy Package 2015 recently adopted by the EC in 28.02.1015 as the Framework Strategy for a Resilient Energy Union with a Forward-Looking Climate change policy highlights the central importance of the research and innovation, recognizes the contribution of the nuclear energy in the total EU’s electricity production, and the need to maintain technological leadership in the nuclear domain, especially in the world's safest nuclear generation.This new generation of nuclear systems (Generation IV) will ensure not only an enhanced safety, but also a more efficient use of the natural resources, which will make them available from hundreds to thousands of years, and a significant reduction of the waste volume and radio-toxicity.Lead-cooled Fast Reactors (LFRs) are numbered among the candidate options by the Generation-IV International Forum (GIF) thanks to the favourable features of lead as coolant, allowing the fulfilment of all the goals of increased safety and reliability, higher sustainability, economic competitiveness and possible more robust proliferation resistance and physical protection. Accordingly, the LFR has been also selected by the European Sustainable Nuclear Industrial Initiative (ESNII), following the mandate – formulated by the Sustainable Nuclear Energy Technology Platform (SNETP) – to support the development of innovative Nuclear Energy System (NES) that will represent a real breakthrough improvement for the sustainability of the nuclear option in the long term. In this context, the FP7 ARCADIA project focuses its efforts to provide a twofold support to the further development of nuclear research programs in the New Member States (NMS), targeting two major areas included in the Strategic Research and Innovation Agenda of SNETP: ESNII, through the support of the ALFRED project towards its realization in Romania, and NUGENIA, approaching remaining safety aspects of Gen III/III+ that could be built in Lithuania, Poland, Czech Republic and Slovenia. On one hand, it focuses on the identification of the primary needs for the ALFRED project and Gen III/III+ reactors, mainly to what concerns supporting Infrastructures and Regulatory aspects (and integrating – for the R&D needs – the outcomes of other research projects in a common frame of National and Regional needs), as well E&T aspects; on the other hand, it investigates the existing National and Regional supporting structures – with a particular attention to the ones in Romania and in all the participating NMS – for defining a map of

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competences potentially eligible to satisfy the previously identified needs. The paper presents in brief the structure and the expectations of the ARCADIA project as well as the most important achievements in the assessment of the competence and infrastructure needed to implement ALFRED in Romania, in defining siting and licensing process, in consolidating an approach for drafting a feasibility study, as well as in reaching the necessary support at national, regional and European level.

111

Transposition challenges of new WENRA requirements into Slovenian regulation

Siniša CimešaUpravaRepublikeSlovenijezajedrskovarnost,Litostrojskaulica54,1000Ljubljana,[email protected]

Learning from events and operating experience is well established process in the nuclear industry. Events almost always disclose some drawbacks in design and operating practice, thus the experience feedback is beneficial for safety. Catastrophic events such as accident in Fukushima, earlier Chernobyl and Three Mile Island have initiated worldwide re-thinking process about the used concepts and approaches, safety re-evaluations and ultimately even in regulations. After the Fukushima accident European countries decided to perform, in very short time after the event, the so called ENSREG Stress tests which ended with National action plans of safety upgrades. WENRA played the important role in this process. Furthermore WENRA also updated the well known WENRA Safety Reference Levels with the objective to support the national authorities in the process of further upgrading of safety in NPP's. The updated WENRA Safety Reference Levels, besides upgrading old WENRA requirements, also give new, key requirements in various technical areas, among which most important are the areas of Design Extension of Existing Reactors and Natural Hazards.In Slovenia we are faced with the challenge to transpose the new post-Fukushima WENRA requirements into domestic regulation. These new requirements shall represent additional motivation for safety improvements, thus the transposition shall be done with care as the new requirements require the period of transition. The paper will describe the process of new WENRA requirements transposition emphasizing the implementation challenges in the real situation at the plant.

112

Quality Assurance System in Nuclear Training CentreTomaž SkobeJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

The paper will present experience from using quality assurance system in nuclear technology and radiation protection courses at Nuclear Training Centre Ljubljana. Nuclear Training Centre has been certified according to ISO 9001:2008 quality standard since December 2006. There are two types of important nuclear technology courses, which we conduct for NPP Krško staff and other organizations, dealing with nuclear technology. The first course is called NPP Technology (the acronym in Slovenian language is TJE) and is intended for future control room operators and the second course, Basic of Nuclear Technology (in Slovenian OTJE) is suitable for other NPP technical personnel, for technical support organizations and

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regulatory bodies. Jožef Stefan Institute is also an authorised institution in the field of Radiation protection and Radiation protection training in Slovenia and a wide spectrum of courses for different users are regularly organized. Paper will present all important steps in course organization, materials preparation, preparation and supervision of lectures and feedback from participants.

113

Main European Union Citizens’ Attitude Influencers with Respect to Nuclear Energy

Pavel Gabriel LazaroUniversityPolitehnicaofBucharest,312,SplaiulIndependentei,Bucharest,[email protected]

Education, training and information to the public are key factors in the governance of risk perception. The main actors responsible for the information sent today to the public about nuclear energy and associated risks are represented by nuclear operators, institutions (state owned or private), Education, Training and Research entities and media. In 2011 the Fukushima accident proved that public felt over-fed with information but trustworthiness of sources was debatable, raising thus the need for further communication improvement.. The opinion of EU citizens on ionizing radiation follows a general pattern, and this pattern depends on level of knowledge, gender, age, whether or not the citizen lives in a country owning a nuclear power plant etc. Trustworthy sources of information are also an element that needs to be analyzed. Main risks associated to nuclear energy are, in general, perceived the same across EU and it is an issue that also needs to be tackled. Trust in information about radiation and its applications depends on a variety of factors such as age, gender, level of education etc.There are different channels the population not involved in nuclear industry receives information through. The emitters of this information are most often mass-media, industry, official agencies and academia. Each source of information is perceived differently by different categories of population (young/old, male/female, graduate/non-graduate etc.). The objective of the paper is to analyze the sources of information for nuclear industry, the senders of such information, at European level, with respect to receivers of the information, the general public. It is expected to identify general public concerns on opinion making factors and to find common characteristics on attitudes from general population across EU. The analysis of the satisfaction degree of the population, regarding the credibility as well as accessibility of information, should be considered in conjunction with existing fears and perceived risks, and should reside in increasing confidence. The sources of information population uses are important to be highlighted, which of these sources are recognized as trustworthy, and what information did the citizens perceive as missing in the communication process related to risks nuclear industry involves.Keywords: ionizing radiation, public perception, information sources, trustworthiness, patterns.

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114

Analysis of radiation protection training results since adoption of new regulation

Jure Hribar, Luka Tavčar, Matjaž KoželjJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

In Slovenia, the law requires that all workers dealing with sources of ionising radiation (“exposed workers”) receive training in radiation protection. Nuclear Training Centre ICJT which is part of the Jožef Stefan Institute is one of the two institutions in Slovenia that is authorized to perform such training. The regulation that specifies this training (“Rules on the obligations of the person carrying out a radiation practice and person possessing a ionizing radiation source” – SV8) was adopted in 2004. For most categories of exposed workers, a 3-day (24 hours or 20 hours) initial training and exam is required. Exposed workers are also required to renew their licence every 5 years by passing an exam. The regulation SV8 does not require any training or lectures before the renewal exam. Nevertheless, we practice a one-day (8 hours) refresher course before this renewal exam.In the past 11 years since adoption of the SV8, about 1,100 exposed workers in industry, science and medicine were trained in ICJT. There were about 450 participants in the initial courses and about 650 participants in the refresher courses. The aim of this paper is to analyse and compare the results of exams in initial and refresher courses, thereby to evaluate the efficiency of short refresher courses.

115

Nuclear Power Infrastructure Assessment and Regional Cooperation in Macedonia


Over the past decades nuclear energy has been proven as reliable and economical energy supply that is capable of meeting demanding energy market requirements. Many countries around the world consider entering into new nuclear energy programs and building new power reactors for satisfying their increasing electrical energy needs. Entering into a new nuclear power program is a major undertaking requiring careful planning and preparation. Preparations for making a decision to enter into a new nuclear energy program requires a significant amount of financial and human resources, time, and assistance from developed countries and international nuclear organizations.The International Atomic Energy Agency (IAEA) from Vienna provides technical help, financial assistance, and documented knowledge that are important for countries facing the challenge of entering nuclear programs for the first time. The IAEA organizes technical courses and information exchange meetings for new countries at which experiences and lessons learned are provided to new countries.For the South-East region in Europe, where countries are fragmented with limited human and financial resources, it is essential to establish good regional cooperation in development and utilization of nuclear energy. Building on some examples of already establish cooperation, recently, supported by the IAEA, a more active regional cooperation program has been initiated, in particular by the Republic of Macedonia. This regional cooperative program is aimed at initiating and maintaining cooperation between countries in the region in making decisions to continue using nuclear power or to enter into a new nuclear power program.

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This paper describes the key activities in the process for making a decision to enter a new nuclear energy program. It describes the efforts currently being conducted in the Republic of Macedonia in the direction of collecting information, performing various feasibility studies, and engaging in regional cooperation for utilizing experiences of the regional countries in performing such activities, and in developing their nuclear power programs.

116

Planning of Energy Demand in Macedonia Using the MAED Model


Over the past decades nuclear energy has been proven as reliable and economical energy supply that is capable of meeting demanding energy market requirements. Many countries around the world consider entering into new nuclear energy programs and building new power reactors for satisfying their increasing electrical energy needs. Entering into a new nuclear power program is a major undertaking requiring careful planning and preparation. Preparations for making a decision to enter into a new nuclear energy program requires a significant amount of financial and human resources, time, and assistance from developed countries and international nuclear organizations.Planning of energy demand in a country is a very important step in developing a strategy for energy sources development, and for coping with the challenges, uncertainties and needs in energy supply. The energy planning is a part of development of energy strategy in a country, and for making a decision to enter a new nuclear power program or to expand an existing program.Several energy demand and supply studies have been performed in Macedonia in the recent years in support of development of national strategy for energy supply. During 2014, as part of the national technical cooperation program with the International Atomic Energy Agency (IAEA), the Macedonian Power Plants (ELEM) has performed a study of the energy demand until 2040 using the IAEA methodology MAED (Model for Assessment of Energy Demand). The study is continuing in 2015 and is focused on the assessment of electrical energy supply.This paper provides a brief description of the technical cooperation program with the IAEA, and a description of the MAED methodology. The paper also provides a brief overview of the energy situation in Macedonia, gives a summary of the energy supply options, and covers the work scope of ELEM, and in particular of the Department for Development and Investments. Finally, the paper discusses the results obtained with the MAED methodology of the energy demand situation in Macedonia until 2040, and compares it with the energy demand in the regional countries of the South-Eastern Europe.

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Thermal Hydraulics


Thermal-hydraulics201

Application of ISA methodology to a Loss of Normal Feedwater ATWS with TRACE 5.0

Maria-Jose RebolloUniversidadPolitechnicadeMadrid(UPM),Alenza4,28003Madrid,[email protected]

Among the sequences outside of design basis, ATWS sequence is one of the most important. Within these sequences the most limiting in PWR are Loss of a Normal Feedwater and Turbine Trip. These type of sequence have to accomplish the acceptance criterion ASME level C (RCS pressure < 22.16MPa) into 10CFR50.62 rule. Based on the possible uprates in PWR 3-loop, the objective is to perform an uprate analysis by means of the Integrated Safety Assessment (ISA) methodology. The impact of several parameters in ATWS sequences has been analyzed, such as power uprate, moderator temperature coefficient and actuation times of AMSAC (turbine trip and AFW injection). The ISA methodology, developed by the Spanish Nuclear Safety Council (CSN) allows obtaining the damage domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest.

202

Uncertainty Quantification of NEPTUNE_CFD calculation by Optimal Statistical Estimator Method

Andrej Prošek, Boštjan Končar, Matjaž LeskovarJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

After revising the rules for emergency core cooling systems (ECCS) in 1988 several uncertainty methods have been developed to evaluate the uncertainty of best-estimate thermalhydraulic system code calculations. The increase of computational power along with continuous development of local mechanistic models opens the space for detailed Computational Fluid Dynamics (CFD) codes in nuclear reactor applications. Unlike the mature uncertainty evaluation of system codes, the uncertainty evaluation for CFD code calculations is still under development.The main purpose of this study was the uncertainty quantification of the NEPTUNE_CFD calculation of the Generic Mixing eXperiment (GEMIX). The same uncertainty approach was used as for the uncertainty evaluation of the large and small break loss of coolant accident using the RELAP5 code. According to the ECCS rule the 95th percentile of the probability distribution is sufficient. The way we choose for the underlying probability distribution function establishment requires the generation of the response surface. The proven Optimal Statistical Estimator (OSE) method has been used for the response surface generation of the NEPTUNE_CFD predictions, which replaces the code calculations when using the Monte Carlo method to randomly sample the input parameters to quantify the 95 percentile of the probability distribution function, accepted as indication that the uncertainties have been accounted for. The turbulent mixing experiment GEMIX performed at the Paul Scherrer Institute was used as a benchmark case. In the GEMIX experiment two turbulent horizontal channel flows with the same fluid properties and inlet velocities have been mixed together to form a mixing layer flow.

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Thermal Hydraulics


Two uncertain input parameters were considered as uncertainty variables: inlet velocity profile and inlet turbulence intensity. The full calculational matrix consisted of 30 NEPTUNE_CFD calculations. The parameter turbulence kinetic energy k was varied in equidistant steps of 1% and the parameter alpha that describes how much the flow is developed (0 ≤ α ≤ 1) was varied in steps of 0.2.The results of the response surface generation together with the uncertainty quantification will be presented. The influence of the number of calculations on the response generation and the uncertainty will be also shown. The obtained uncertainty results suggest that OSE is very powerful for uncertainty evaluation of CFD calculations (especially when few input parameters are varied) due to its ability to model highly nonlinear functions.

204

Accurate wall-resolved Large Eddy Simulation of a turbulent flow in 5×5 fuel rod bundle

Blaž Mikuž, Iztok TiseljJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Efficiency and safety of a pressurized water reactor during normal operation depends on the heat removal capability of the turbulent flow in a reactor core. Heat transfer from the surface of the fuel rods to the surrounding water in reactor core is enhanced with efficient mixing of the coolant flow. For this reason, the fuel assemblies contain passive mixing devices, i.e. mixing vanes, which induce additional cross flow in the subchannels of the fuel assemblies. As a result, directly downstream of the mixing vanes, the mixing of the flow is strong and governed by advection. However, further downstream the cross flow is being damped until it dies out. Here, the turbulent mixing is characterized by the anisotropy of turbulent fluctuations, which can become a predominant turbulent mixing process and induces the formation of the secondary flow structures due to Prandtl’s 2nd mechanism. This phenomena is weak, however it is believed to be crucial for correct prediction of the turbulent mixing in a bare (undisturbed) fuel bundle.In the present paper, the wall-adapting local eddy-viscosity (WALE) model has been implemented in OpenFOAM code and applied for wall-resolved Large Eddy Simulation (LES) of a turbulent flow in 5×5 fuel rod bundle at moderately high Reynolds number, i.e. Re=50000. The numerical model is validated against Laser Doppler Velocimetry (LDV) measurements of the MATiS-H experiment, which was performed in 2011 at KAERI, South Korea. The velocity profiles are in a good agreement with measurements since most of them are within the measurement error. Our numerical approach is computationally demanding since the near-wall region is resolved, however it provides us also accurate velocity fields in the near-wall regions, which were not measured due to the measurement technique limitations. The shape of the velocity fluctuations is correctly captured while the level of the resolved fluctuations is underpredicted for about 25 percent. A slight shift of the calculated velocity extremes towards the outer walls of the fuel bundle is detected, which is consistent with observations of some other authors, who used wall-modelled LES and Reynolds Stress models.

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Thermal Hydraulics


205

Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel

ExperimentsSergii Lutsanych1, Fabio Moretti2, Francesco Saverio D'Auria3

1 UniversityofPisa,ViaMontebellodiMezzo17,19020Bolano,Italy2 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,Italy3 UniversityofPisa,SanPieroaGradoNuclearResearchGroup(GRNSPG),ViaLivornese1291, 56122Pisa,[email protected]

This paper deals with the validation of the multifield computational fluid dynamics code NEPTUNE_CFD 2.0.1 against experimental data available from the OECD/NRC NUPEC PWR subchannel and bundle tests (PSBTs) international benchmark. The present work is performed in the framework of the NURESAFE European collaborative project and focuses on the steady-state single subchannel void fraction tests.12 different experimental runs (out of a total 126) covering a wide range of test conditions have been selected and simulated for a test section representing a central PWR fuel assembly. Following the NEA/CSNI best practice guidelines about computational grid design and grid quality, mesh sensitivity analysis has been performed using axial and radial grid refinement. Both axial and radial mesh sensitivity studies do not exhibit any significant change in the predicted results. Besides, a series of sensitivity calculations have been performed in order to investigate the influence of uncertainties of the experimental boundary and initial conditions on the code predictions. The influence of code physical and closure models on the void fraction prediction has been studied and discussed in detail. Generally, the calculated cross-sectional averaged void fraction at the measurement plane differs from the measured one by maximum of +/- 8%. This discrepancy is comparable to the 2-sigma experimental uncertainty range on void fraction measurement. The performed investigations have shown the ability of NEPTUNE_CFD to predict reasonably the void fraction in PSBT subchannel using appropriate modelling.

206

Numerical Investigation to Examine Dust impacts on the Dry Cask passive Cooling under U.A.E. harsh environmental

conditionsYacine Addad, Alia Mohamed Ahmed Hammad AlghafriKhalifaUniversityofScienceTechnologyandResearch,PO.Box127788,AbuDhabi,[email protected]

The United Arab Emirates (UAE) is on its way to commission the first of four planned nuclear power plants by May, 2017. This first reactor is expected to start its first discharge of spent nuclear fuel by 2018 or 2019 depending on its refueling schedule. The current plan is to first store the spent fuel in a water filled pool. However, in order to store spent fuel generated beyond 20 years up to 60 year plants life, an additional mean has to be provided. The common solution, adopted by a number of countries, is to use the dry storage concept. This solution presents some challenges. In addition to the costs and technical issues discussed in the paper presented by Al Saadi et al. [1], local issues that have to be also taken in consideration are related to harsh environmental conditions specific to the Gulf Cooperative Council

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(GCC) region. These are the relatively high temperatures during the whole year, regular sand storms, and a nearly permanent presence of dust in the atmosphere. Furthermore, Sabkha (or Evaporates) ground soil is site-specific characteristics of the region of the NPP under construction and has to be also taken in consideration. As the spent fuel storage is a passive heat dissipation system, it will be strongly dependent on local environmental conditions, i.e., the thermodynamic properties of the air flowing into the Ventilated Concrete Cask (VCC). The air high temperatures, especially during summer time, would mean that the temperature difference between heat sink, air, and the spent fuel (heat source) is small. As the flow in the ventilated concrete cask is driven by the buoyancy term which on its turn is dependent on this temperature difference, small temperature differences would mean low heat removal, especially if the flow regime is also affect (from turbulent to laminar). The dust particles suspended in the air (natural nanofluids) are expected to change the air thermodynamic properties, in particular, the density (impact on buoyancy forces) and conductivity (impact on thermal resistance), hence; the objective of this study is to investigate, using CFD numerical study, whether these changes in the properties are to improve or further deteriorate the heat transfer efficiency in the dry storage. Beforehand, a validation test case is conducted to examine the CFD code (Star-CCM+) and the turbulence models capabilities to predict such two-phase low-level turbulent flow with conjugate heat transfer.

207

NPP Krško DVI LOCA Calculation Using RELAP5/mod 3.3 and FRAPTRAN to Assess UFC Modification Influence

Vesna Benčik1, Davor Grgić1, Siniša Šadek1, Nikola Čavlina2

1 UniversityofZagreb,FacultyofElectricalEngineeringandComputing,Unska3,10000Zagreb,Croatia2 FakultetelektrotehnikeiračunarstvaZagreb,Unska3,10000ZAGREB,[email protected]

Damaged fuel assemblies have been identified at NPP Krško during 2013 outage refueling activities. As a potential root cause for fuel assemblies located at the core periphery, the baffle jetting across the baffle joints between the baffle-barrel bypass and the core was identified. In order to decrease the pressure difference across the baffle joints the Upflow Conversion (UFC) modification has been performed at NPP Krško during 2015. outage. The modification consisted in altering the reactor vessel internals in such way that the coolant downflow path in the baffle-barrel region was converted to an upflow path. Thereby, the pressure difference across the baffle joints has been minimized and the baffle jetting as a major cause of damage of peripheral fuel elements has been eliminated. The UFC modification has affected the thermal-hydraulic as well as reactor vessel internals and fuel dynamic and structural behavior. The performed safety analyses showed that most of the influence should be related to Small Break Loss of Coolant Accident (SB LOCA) and Large Break (LB) LOCA. As a support to the UFC modification review, thermal-hydraulic analyses of the Direct Vessel Injection (DVI) line SB LOCA have been analyzed using RELAP5/mod3.3 code along with the FRAPTRAN calculation for hot fuel rod. Safety analyses have been performed taking into account conservative assumptions. Both the downflow configuration (before UFC) and the upflow configuration (UFC) were considered in the analyses to assess influence of the modification. Depending on the assumptions used in calculation small increase in Peak Cladding Temperatures (PCT) was predicted due to UFC, but safety limits are not endangered.

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208

Turbulent flow simulations of wire-wrapped fuel pin bundle of sodium cooled fast reactor in lattice-Boltzmann framework

Ali Tiftikci, Cemil KocarHacettepeUniversity,NuclearEngineeringDepartment,06800Beytepe,Ankara,[email protected]

Lattice-Boltzmann method (LBM) simulations were performed for sodium cooled fast breeder type nuclear reactors with helical spacer wires in order to validate the Smagorinsky (LES) and Very Large Eddy Simulations (VLES) turbulence models implemented in lattice-Boltzmann framework. The simulations are handled for hexagonal fuel rod bundle for seven rods with one helical pitch length geometry. For each turbulence model study, post-processes turbulence quantities such as velocity profiles and Reynolds stresses are compared. Additionally, the friction factor obtained from LES and VLES are compared with the experimental data. The comparisons show that LBM simulations are in good agreement with experimental correlations. Validation of written code in C++ programming language is encouraging for the simulation of more complex geometries for the nuclear reactor fuel bundle system in the future

209

Analysis of Channel Blockage of MNSR Reactor Using the System Thermal-Hydraulic Code RELAP5/MOD3.3

Simon Adu1, Ivan Horvatovic2, Francesco Saverio D'Auria3, Benjamin B.J.B Nyarko1, Emi-Reynolds Geoffrey1, Ofor Darko Emmanuel1

1 GhanaAtomicEnergyCommission,RadiationProtectionInstitute,P.O.BoxLG80,00233Accra,Ghana2 NuclearResearchGroupofSanPieroaGrado,SanPieroaGrado,56122Pisa,Italy3 UniversityofPisa,SanPieroaGradoNuclearResearchGroup(GRNSPG),ViaLivornese1291, 56122Pisa,[email protected]

The increased extensive use of research reactors and improved regulatory and operational safety requirements have increased the use of more realistic simulations of the plant phenomena involved during steady-state and transient conditions. The earlier adopted conservative model assumptions in the reactor safety analysis which were based on conservatism will be replaced with best-estimate methodologies. The best-estimate approach aims at providing a detailed realistic description of postulated accident scenarios based on best-available modeling methodologies and numerical solution strategies sufficiently verified against experimental data from differently scaled separate and integral effect test facilities.The core behavior of Ghana Research Reactor one (GHARR-1) Miniature Neutron Source Reactor (MNSR) during the loss of flow has been investigated. Steady-state and transient analysis were done with best estimate code RELAP5/MOD3.3. The simulated transient characterizes a Loss-of- Flow-Accident (LOFA) type transient. The study forms part of the ongoing core conversion program that is currently ongoing at the facility to convert the reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The partial and total blockage of coolant to the reactor core transients were performed to study the behavior of the reactor. It was observed in the case of partial blockage that although boiling occurred in the blocked channels, which lead to increase in both coolant and cladding temperature, the reactor presented a safer steady again due to in-flow of coolant from adjacent channels to the blocked channels. The calculations showed that cladding and coolant temperatures of blocked channels are below the melting point of the assembly. For total blockage the calculations ended abruptly at about 70s after the start of transient. Therefore we could not observe the whole transient but the observed phenomena indicate unsafe behaviour of the reactor.

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210

Researches Made in Order to Estimate the Implications of Temperature Variations from the Spent Fuel Bay over the

Corrosion Rate of the Spent Fuel Cladding ElementsDiana Laura IcleanuCITON,111Atomstilor,P.O.B.5204-MG-4,76900Bucuresti-Magurele,[email protected]

This paper aims to develop a thermal-hydraulic analysis for the Spent Fuel Bay Cooling and Purification System from a CANDU 6 Nuclear Power Plant. The Spent Fuel Bay Cooling and Purification System provide cooling and purification of the water within the spent fuel storage bay and for three auxiliary (discharge, reception and defective fuel storage) bays. In order to control the corrosion of metal surfaces of both fuel elements sheaths and underwater bay components, a chemical control of the Spent Fuel Bay System has to be performed. However, corrosion processes can’t be arrested and depends on several factors, such as the temperature. The paper will develop an analysis of the potential implications of temperature values variations in the Spent Fuel Bay on the corrosion behavior of fuel elements claddings. Using Flowmaster calculation code, the temperature evolution in the Spent Fuel Bay will be modeled in order to quantify the effects of temperature value modification on the corrosion rate of fuel elements claddings. The Flowmaster calculation code is going to be used to develop models and calculation assumptions, their geometric configuration and, also, to define input data for hydraulic analysis and calculation assumptions or input data for thermal calculation verification operation of heat exchangers that are part of this system. The results obtained after performing the calculations with the Flowmaster code are going to be used as basis for discussions that will be carried in order to formulate the conclusions regarding the potential implications of temperature and heat flux variations on the amplitude of the corrosion phenomena.

211

Simulation of Turbulent Liquid Metal Flow in a Triangular Rod Bundle Sub - Channel

Miha Pogačar1, Ivo Kljenak2, Matej Tekavčič2

1 UniverzavLjubljani,Fakuletazastrojništvo,Aškerčevacesta6,SI-1000Ljubljana,Slovenia2 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Liquid heavy metal thermal-hydraulic behaviour and its effect on the performance of liquid-metal cooled nuclear reactor core bundles have been the subject of research in the past. Using Computational Fluid Dynamics (CFD) software, a reliable prediction of physical phenomena occurring in the reactor core is possible. The acquired results can be subsequently used in designing the reactor core itself. In the first part the work presents the steady-state flow phenomena through a hexagonal fuel rod bundle in the Advanced European Lead Fast Reactor European Demonstration (ALFRED), modelled using the ANSYS Fluent CFD code. A comparison of the axial evolution of the velocity profile and pressure loss between the Renormalization Group (RNG) k-ε model and the Reynolds Stress (RSM) model has been made to show the effect the chosen turbulence model has on the resulting flow-field. A qualitative comparison of the secondary flow pattern has been made. The second part studies the effect of domain discretization on flow convergence through a grid convergence study (GCS).

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212

Preliminary study of the LORELEI test device with the CATHARE-2 code

Paolo Battistoni1, Marco Sumini1, Sandro Manservisi1, Christian Gonnier2, Lionel Ferry2, Didier Tarabelli2

1 UniversityofBologna,FacultadiIngehneria,VialeRisorgimeno2,40136Bologna,Italy2 CEA,MemberofSNETPExecutiveCommittee,GifsurYvette91191,[email protected]

In the Jules Horowitz Reactor under construction in Cadarache, one of the foreseen experimental facilities is the LORELEI (Light water One Rod Equipment for Loca Experimental Investigation) device. The aim of the test device is to analyze the thermomechanical behavior of the cladding and the fuel rod, and the radiological consequences of a Large Break, LOCA (Loss-Of-Coolant Accident) accident One of the aspects of the design studies is to simulate the behavior of the device in normal operation and in incidental and accidental situations, thanks to the CATHARE2 code. The CATHARE2 code is a thermal-hydraulic two-phase flow software aiming at simulating nuclear circuits.In the actual paper, we present two different numerical simulations corresponding to the two main working situations of the LORELEI device.- The first working situation is the re-irradiation phase, when the device is filled with water and the fuel rod cooled in a natural convection loop. The simulations will provide temperature profiles and velocities in the thermosyphon flow, both in single and two phase flow in order to determine power thresholds up to which the thermosyphon will be stopped.- The second working situation, consecutive to the re-irradiation phase, is the LOCA sequence when the device is in a gas environment (steam, helium and hydrogen). For this high temperature situation, a dedicated radiative heat transfer module is implemented. The objective of the simulation is to obtain representative thermal maps of all the structures from the beginning of the LOCA sequence up to the high temperature plateau. Calculations will be compared with CFD calculations.

213

Influence of Liquid Inlet Modeling on Simulated Wave Characteristics in Vertical Gas-Liquid Churn Flow

Matej Tekavčič, Boštjan Končar, Ivo KljenakJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

The churn flow regime in gas-liquid two-phase flow in vertical pipes can be viewed as a transitional regime between slug flow and annular flow and is often related to the onset of flooding phenomena, which is of particular interest for safety analyses of the loss-of-coolant accident in liquid water nuclear reactors. In that regime, large waves of liquid travelling upwards can typically be observed.Wang et al. [1] examined the properties of flooding type waves in the churn flow of air and water in a vertical pipe with 19 and 34 mm internal diameter (i.d.). In the test section of the experimental setup, a transparent porous wall liquid injector with uniformly distributed holes was used, which enabled a smooth liquid entry. A similar experimental setup was used by Barbosa et al. [2] who studied churn flow in a 32 mm i.d. pipe. A comparison of the experimental results from the literature performed by Wang et al. [1] showed that the measured wave frequency is approximately proportional to the gas Reynolds number and to the cross-sectional average of net liquid mass flow rate upwards, with one notable exception in the data set of Hewitt et al. [3] in a 10 mm i.d. pipe.

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Our previous attempts to model such flooding waves in churn flow in a 19 mm i.d. pipe using two-phase CFD, showed an agreement of calculated wave amplitudes with experimental data, but significant disagreements of calculated wave frequencies in certain cases (e.g. gas Reynolds number of 6000). At the same time, simulation results in 32 mm i.d. pipe [4] agree well with the experimental data of Barbosa et al. [2], implying that the pipe geometry has an impact on the wave frequency. In both simulations, the porous wall liquid inlet was modelled as a simple surface liquid inlet of 6 mm height.In the present paper, the impact of the liquid inlet model on the flooding wave frequency was quantified and a sensitivity study was performed. The numerical domain consists of an axis-symmetric wedge with the porous wall inlet region representing a vertical pipe experimental test section. A transient simulation of an isothermal churn flow of air and water was performed using two-phase CFD approach with interface sharpening. The gas and liquid phases are considered immiscible and incompressible with no mass transfer between them. The study is essential to ensure that the characteristics of simulated waves are determined by the modeled physical phenomena of gas and liquid flow and interaction, and not by the modeling of the liquid inlet.References:[1] K. Wang, B. Bai, W. Ma, Chem. Eng. Sci. 104(0), 2013.[2] J.R. Barbosa, A.H. Govan, G.F. Hewitt, Int. J. of Multiphase Flow 27(12), 2001.[3] G.F. Hewitt, C.J. Martin, N.S: Wilkes, Physicochemical Hydrodynamics 6(1/2), 1985.[4] E. Da Riva, D. Del Col, Chem. Eng. Sci., 64(17), 2009.

214

Experimental Investigation on Powder Conductivity for the Application to Double Wall Bayonet Tube Bundle Steam

GeneratorDavide Rozzia1, Giuseppe Fasano2, Mariano Tarantino2, Alessandro Del Nevo2, Nicola Forgione1, Alessandro Alemberti3

1 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,Italy2 ENEACRBrasimone,LocalitaBrasimone,40032Camugnano(BO),Italy3 AnsaldoNuclearS.p.a.,C.soF.M.Perrone25,16152Genova,[email protected]

The 300Mth Advanced Lead cooled Fast Reactor European Demonstrator (ALFRED) adopts the super-heated steam double wall once through bayonet type bundle as SG. The single unit is constituted by three concentric tubes and is based on the concept to provide a double physical separation between the coolant (water) and the hot fluid (lead) in order to increase the safety margin of the plant by reducing the probability of interaction coolant-hot fluid. Furthermore, this configuration allows the possibility to monitor eventual leakages from the coolant or from the hot fluid by pressuring the annular region between the double walls. On the other hand, since its goal is to achieve high thermal performance, the annular space that separates the fluids should be filled with a porous heat transfer enhancer (i.e. Si-C powder, stainless steel powder).In order to fulfill the R&D need related to selection of the porous heat transfer enhancer, the Tubes for Powders (TxP) facility has been designed and constructed during 2012. The facility consists of three concentric tubes and it allows to estimate the conductivity of a porous material by measuring the temperature drop across its borders and the removed power. It has the capability to introduce helium and to pressurize it up to 5 bar. The TxP facility has been operated for about one year with the purpose to assess the conductivity of powders both in air environment and in pressurized helium atmosphere.The present paper focuses on the experimental campaigns carried out to investigate the thermal performance of AISI-316 powders in support to the design of the HERO test section (which is presently

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under commissioning and consist of a bundle of 7 tubes representing the ALFRED SG). Twelve experimental campaigns will be presented in order to characterize the conductivity of AISI-316. The first matrix of tests aims to assess the effect of cyclical thermal loads and is conducted in air environment. Once determined the cyclical behavior, the conditioned powder, has been tested under helium environment at four different levels of pressurization.

215

Steady-State Calculation of Krško NPP TRACE model with Three Dimensional Pressure Vessel

Ovidiu-Adrian Berar, Andrej Prošek, Borut MavkoJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

In recent years, the need for multidimensional prediction of thermal-hydraulic conditions in the case of some Design Basis Accident analysis has become evident. TRACE is an advanced, best-estimate reactor systems code capable of solving the fluid-dynamics equations in three dimensional space by using a specialized VESSEL component. The objective of the current work was to study the TRACE vessel model nodalization influence on the steady-state calculation results. The Krško RPV TRACE model was developed using the VESSEL component employing a cylindrical geometry model discretization in axial levels, radial rings, and azimuthal sectors. Specific physical geometry of RPV such as the downcomer, lower plenum, core section, and upper plenum was used for the three-dimensional RPV modeling. The fuel assemblies are modeled using Heat-Structures connected to the VESSEL component. The three-dimensional model of the RPV has been separately tested prior to incorporation into the one-dimensional plant model. The one-dimensional TRACE input model of Krško NPP has been obtained from an existing RELAP5 input deck through semiautomatic conversion and manual input. A steady-state calculation of the Krško NPP model using the three-dimensional Reactor Pressure Vessel (RPV) model was performed using the TRACE V5.0 Patch 4 code and the results are compared to the RELAP5 steady-state calculation. A sensitivity study assessing the influence of the RPV nodalization schemes on the predicted steady-state conditions will be presented and discussed.

217

Effect of the mass flow rate and the subcooling temperature on pressure drop oscillations in a horizontal pipe

Il Woong Park, Maria Fernandino, Carlos DoraoNorwegianUniversityofScienceandTechnologyDepartmentofChemistry,NO-7491Trondheim,[email protected]

Pressure Drop Oscillation (PDO) is a particular case of two-phase flow instabilities (1). A PDO occurs in a system having compressible volume upstream or within the heated section and when the system operates in the negative slope region of the N-shape curve, i.e. pressure drop vs. flow rate curve. PDOs have a long period (of the order of dozens of seconds) and produce big excursions of the flow resulting in large variations in the local wall temperature (thermal oscillation). Previous studies have shown that mass flux and inlet subcooling influence the period and amplitude of the PDO. However, different studies have shown different trends (2-4). The objective of this study is to investigate the effect of the mass flux and inlet subcooling in the characteristics of the oscillations. The liquid level and the pressure of the surge tank were controlled, to

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prevent the impact of them on the PDO. The plot of the pressure drop along the mass flux of heated pipe with varying subcooling temperature was obtained to find the region where the PDO occurs. It was discovered that periods and amplitudes of the PDO were not linear relation of the mass flow. They were increased with increasing the mass flow rate until certain points and decreased after that point.

(1) Chiapero, E. Manavela, M. Fernandino, and C. A. Dorao. »Review on pressure drop oscillations in boiling systems.« Nuclear Engineering and Design 250 (2012): 436-447.(2) Yüncü, H., O. T. Yildirim, and S. Kakac. »Two-phase flow instabilities in a horizontal single boiling channel.« Applied Scientific Research 48.1 (1991): 83-104.(3) Çomakli, Ö., S. Karsli, and M. Yilmaz. »Experimental investigation of two phase flow instabilities in a horizontal in-tube boiling system.« Energy Conversion and Management 43.2 (2002): 249-268.APA(4) Ding, Y., S. Kakac, and X. J. Chen. »Dynamic instabilities of boiling two-phase flow in a single horizontal channel.« Experimental Thermal and Fluid Science 11.4 (1995): 327-342.

218

Downcomer boiling phenomena analysis during large break loss of coolant accident in APR1400

Safa Mohamed Aidaroos Salem Alhashmi, Ho Joon Yoon, Yacine AddadKhalifaUniversityofScienceTechnologyandResearch,PO.Box127788,AbuDhabi,[email protected]

The Blowdown is one of the research areas of interest to the Emirates Nuclear Energy Corporation (ENEC) in the United Arab Emirates. The investigation and the assessment of these complex phenomena occurring at the blowdown and reflood periods either as separate effects or as aggregate contribution to the integral system are of primordial importance to the nuclear power plants safety studies. In particular, the ability of system codes (such as RELAP, MARS, and TRACE) to correctly reproduce the coolant behavior and thermodynamic properties under these conditions has to be continuously validated and improved [Roth et al. 2014]. Examples of similar studies can be found in [Yun et al. 2008], [Freixa et al. 2011], [Huh et al. 2006], [Joon et al. 2002] and [Haejung et al. 2012].Hence, the aim of the present study is to conduct a numerical study using the system codes (TRACE and RELAP) to predict the Downcomer boiling phenomena occurring during the reflood phase of a large-break LOCA for the APR1400 and compare with the experimental data obtained from the Downcomer Boiling (DOBO) test facility reported in [Yun et al. 2008]. Although this test case has been already studied using the system codes mentioned above, it is revisited herein using the latest versions of the codes to examine the effect due to the codes recent modifications in terms of correlations and their implementation and to understand the main factors affecting the discrepancies between the codes predictions observed in the previous studies. It worth mentioning here that the initial results obtained in the present work, to be reported in the full version of the paper, are much superior to previous ones reported in the open literature. The reasons behind these predictions enhancements are under investigation and the results analysis in progress will be also reported in the paper.

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219

Spectral element direct numerical simulation of heat transfer in turbulent channel sodium flow

Jure Oder, Iztok TiseljJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Phénix was a prototype fast breeder reactor that was in operation between the year 1973 and 2009. It was a pool-type reactor cooled with liquid sodium. The net power generating capacity was around 230 MW and the breeding ratio of about 1.12.An inspection of the Phénix reactor in May 1993 discovered cracks in some of its equipment. The cracks were found in a sodium discharge area where the sodium from the hot leg (550 °C) of the loop flows into the expansion tanks operating at cold temperature (350 °C). Metallurgical observations confirmed that the defects were induced by high thermal fatigue. Because of high thermal conductivity of sodium and liquid metals in general, thermal fluctuations in the liquid can penetrate into adjacent structures with low attenuation. Due to turbulent nature of the flow the thermal fluctuations at some point have a typical frequency of 1 Hz. Together with high thermal conductivity these thermal loads cause quick ageing of materials.In this paper we present the direct numerical simulations of fully developed turbulent flow in a channel between two plates with finite dimensions. The outer walls of solid plates are heated with constant flux. In the two other directions periodic boundary conditions are used. To thermally couple the walls with the fluid, conjugate heat transfer model is used. To compare the results to the past results, gravity is neglected and the temperature is a passive scalar.Simulations are performed with the NEK5000 code. The most notable feature of this code is the use of spectral elements to solve for velocity, temperature and any other passive scalar.This work is part of work that is performed within the SESAME project of Horizon2020 research programme and is a continuation of research at our department. The main purpose of this work is to compare the results obtained with NEK5000 to justify its use in future simulations of more complex geometries.

220

Modeling of NEK Containment in computer code APROSJure JazbinšekZEL-ENrazvojnicenterenergetike,Hočevarjevtrg1,8270Krško,[email protected]

Nuclear power plant Krško (NEK) reactor containment building nodalization in computer code APROS is developed based on available plant’s documents and Krško NPP GOTHIC nodalization notebook. The free volume calculation, calculation of concrete volume and surface area, and calculation of the volume displaced by main components was performed. The heat structures data, except for interior concrete that is explicitly calculated, is based on NEK Updated Safety Assessment Report (USAR) Chapter 6 passive heat structures. Following a postulated rupture (double ended hot leg (DEHL) loss of coolant accident (LOCA)) of the Reactor Coolant System (RCS), steam and water is released into the containment atmosphere. When the break occurs, the water passes through the break where a portion flashes to steam at the lower pressure of the containment. These releases continue until the RCS depressurizes to the pressure in the containment (end of blowdown). The analysis assumes that the lower plenum is filled with saturated water at the end of blowdown, to maximize steam releases to the containment. Simulation results of pressure and temperature in containment, surface temperatures and water volume in the sumps representing freshly condensed water from blowdown event will be presented.

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221

FONESYS and SILENCE Networks: Looking to the Future of T-H Code Development and Experimentation

Sergii Lutsanych1, Fabio Moretti2, Nusret Aksan3, Francesco Saverio D'Auria4, Alessandro Petruzzi4

1 UniversityofPisa,ViaMontebellodiMezzo17,19020Bolano,Italy2 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,Italy3 NuclearResearchGroupofSanPieroaGrado,SanPieroaGrado,56122Pisa,Italy4 UniversityofPisa,SanPieroaGradoNuclearResearchGroup(GRNSPG),ViaLivornese1291, 56122Pisa,[email protected]

The purpose of this paper is to present briefly the projects called FONESYS (Forum & Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics) and SILENCE (Significant Light and Heavy Water Reactor Thermal Hydraulic Experiments Network for the Consistent Exploitation of the Data), their participants, the motivation for the projects, their main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor technology, safety and design. Namely, the number of code-users can easily be predicted to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework the FONESYS initiative has been launched in 2010. The main targets of FONESYS are:• To promote the use of SYS-TH Codes and the application of the BEPU approaches;• To establish acceptable and recognized procedures and thresholds for Verification and Validation (V&V);• To create a common ground for discussing envisaged improvements in various areas, including user-

interface, and the connection with other numerical tools, including Computational Fluid Dynamics (CFD) Codes.

According to FONESYS statute, there are seven signatories Institutions and two observer Institutions currently participating in the project. Signatories Institutions are AREVA NP SAS (AREVA-NP), Commissariat a l’Énergie Atomique et aux Énergies Alternatives (CEA), San Piero a Grado Nuclear Research Group - University of Pisa (GRNSPG-UNIPI), Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Korea Atomic Energy Research Institute (KAERI), Korea Institute of Nuclear Safety (KINS), and VTT Technical Research Centre of Finland. SILENCE is a network that intends to promote the cooperation among teams of experimentalists managing or involved in significant experimental projects in nuclear reactor thermal-hydraulics, with the aim to contrast the risk of losing expertise and vision in this important area of the nuclear technology. This network was launched in 2012, replicating for the TH experimental domain the role that FONESYS plays in the code-development domain. Currently, are Members of SILENCE the following Organizations: AREVA GmbH, Helmholtz Zentrum Dresden-Rossendorf (HZDR), Korea Atomic Energy Research Institute (KAERI), Hungarian Academy of Sciences Centre for Energy Research (MTA EK), Lappeenranta University of Technology (LUT), Oregon State University (OSU) and Paul Scherrer Institute (PSI). SILENCE is currently organizing a “Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics” (SWINTH-2016).The San Piero a Grado Nuclear Research Group - University of Pisa (GRNSPG-UNIPI) is the Host Institution and plays as a Scientific Secretariat for both projects.

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222

Determination of Geometrical and Operating Parameters of PRHR for VVER Reactors: Cooling by Natural Circulation of

Atmospheric AirHüseyin Ayhan, Cemal Niyazi SökmenHacettepeUniversity,NuclearEngineeringDepartment,06800Beytepe,Ankara,[email protected]

Passive safety concept was proposed to improve safety and reliability of nuclear power plants. However, for the nuclear technology, passive systems require a special attention, since the disadvantages for their use come from the larger difficulties in the thermal-hydraulic design compared with active systems. One of the strong dynamics of the working principle of passive systems is natural circulation. Natural circulation phenomena play an important role in energy transfer from hot zones to the cold zones without using a mechanical pump. In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In this study, the natural circulation phenomena are studied with reference to station blackout scenario in VVER type Nuclear Power Plant (NPP). Analytical calculations are considered to evaluate the natural circulation performance of the VVER type NPP. Thermal-hydraulic calculations of Passive Residual Heat Removal System (PRHRS) were performed. For many geometrical combinations and boundary conditions were investigated.

Keywords: Natural Circulation, RHR Systems, VVER Reactors

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Research Reactors301

Human Resource Development and Nuclear Education Through Research Reactors: Successful Approach to Build Up

the Future Generation of Nuclear ProfessionalsHelmuth Böck1, Mario Villa1, Andrea Borio Di Tigliole2, Judy Vyshniauskas2, Lubomir Sklenka3, Luka Snoj4, Attila Tormási5

1 TechnicalUniversityVienna,Atominstitut,Stadionallee2,1020Vienna–Austria2 ResearchReactorSection,DepartmentofNuclearEnergy,InternationalAtomicEnergyAgency, WagramerStreet,1220Vienna–Austria3 DepartmentofNuclearReactor,CzechTechnicalUniversityinPrague,VHolesovickach2, 18000Prague8–CzechRepublic4 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia5 InstituteofNuclearTechniques,BudapestUniversityofTechnologyandEconomicsMuegyetemrkp.9,[email protected]

Research reactors around the world, with their diverse designs, power level and utilization profiles have played and continue to play an important role in the development of nuclear science and technology with very specific applicability in various fields. Besides playing an important role in the medical, industrial and material research fields, Research reactors have proven to be a useful tool for nuclear human resource development, training and education of the future generation of nuclear engineers and nuclear scientists. This paper will give an overview of the utilization of RRs for education and human resources development within the nuclear science and technology field. It will elaborate on the challenges of the research reactor (RR) community and will present the efforts of the International Atomic Energy Agency (IAEA) to enhance the utilization of RRs in developing countries via the application of nuclear education and public outreach that contributes to the early development of human resources. Emphasis will be given to RR coalitions and networks in particular the Eastern European Research Reactor Initiative (EERRI) coalition and the tangible results that this coalition has been delivering over the years since 2009. The coalition will be introduced and the dynamics of the host institutions will be explained. The coalition has been organizing a joint training course, the EERRI Group Fellowship Training Programme. This course will be portrayed as a successful example of international RR collaboration that has helped capped some of the human resources challenges faced today by institutions and their governments in meeting nuclear education requirements, by training around 80 international students, coming from diverse countries, seeking for hands-on training at nuclear facilities. The statistics of the participants to the EERRI Group Fellowship Training Programme will be analysed and the advantages of such strategy of the IAEA to continue expanding such approach to other regions will be describe. Finally this type of collaboration will be portrait as a potential activity to increase utilization of RRs and more importantly as an excellent example of why RRs are needed in the future to encourage young people to follow a path in nuclear science and technology.

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302

Measurements with Multiple In-core Fission Chambers at the JSI TRIGA Mark II Reactor

Tanja Kaiba, Gašper Žerovnik, Luka SnojJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

The aim of the collaboration between the CEA Cadarache and the JSI was to improve the accuracy of current online power monitoring system at the JSI TRIGA reactor. In small research reactors such as the JSI TRIGA, the neutron flux profile inside the reactor core changes significantly in the axial and radial direction. Moreover, the effect of the neutron flux redistribution due to the control rod movements is significant. Until now the reactor ex-core power monitoring system has been used, which is highly sensitive to the effects mentioned above. Therefore a new online in-core system has been devised which would utilize multiple miniature detectors distributed throughout the core to average out the dependence of the signal on the detector and control rod position. The measurements of the fission rate axial distribution using the CEA developed fission chambers inside different measuring positions simultaneously, were performed at the JSI TRIGA research reactor. The measurements were compared with the calculated axial fission rate profiles by the Monte Carlo neutron transport calculations with the MCNP code. In addition a radial neutron flux and fission rate profiles were calculated at different axial positions. To verify the calculations, they were compared with the measurements at the different control rod positions. In general agreement between the relative measurements and simulations is good, which confirms the accuracy of the existing MCNP model and its use for the neutron flux redistribution evaluation. Furthermore, using multiple fission chambers inside the reactor core can average out and minimize the effects of the neutron flux redistribution on the reactor power measurements.

303

Loss of flow accident analyses in Tehran Research ReactorAhmad LashkariAtomicEnergyOrganizationofIran,KHAfrica,K.Tandis,No.7P.O.Box14155-1339,19156Tehran,[email protected]

In recent decades, there has been a concerted effort to develop simpler computational tools for simulating thermal-hydraulic behavior of reactor system during real and hypothetical transient scenarios. In the recent years an increasing trend in applying what may be called more fundamental modeling approaches to elucidate details of the reactor dynamic performance has been seen in the literature and recognized as very useful by the nuclear industry. In all previous works, rarely the results of these models were compared with the experimental values and usually compared with the results of other codes. Experimental and numerical studies of loss of flow accident in TRR are the main objective of this work. In this paper in addition to reporting the experimental results of loss of flow experiment, a numerical model is presented and used to analyze a series of loss of flow transients in TRR. The model predictions are compared with the experiment and PARET code results. The model uses the piecewise constant and lumped parameter methods for the coupled point kinetics and thermal-hydraulics modules respectively. The advantages of the piecewise constant method are simplicity, efficient and accurately. A main criterion on the applicability range of this model is that the exit coolant temperature remains below the saturation temperature, i.e. no bulk boiling occurs in the core. The calculation values of power and coolant temperature in loss of flow scenario’s, are in good agreement with the experiment values. However, the model is a useful tool for the transient analysis of most research reactor encountered in practice. The main objective of this

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work is using simple calculation methods and benchmarking them with experimental data. Also the inlet temperature of the coolant is not constant as an input parameter (unlike PARET code). This model can be used for research and training proposes.

304

Analysis of coolant temperature distribution for the validation of TRIGA Mark II CFD computational model

Romain Henry, Marko MatkovičJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

TRIGA Mark II reactor at the »Jožef Stefan« Institute in Ljubljana is a 250 kW pool type reactor, which is used for Neutron Activation Analysis, Neutron Radiography and Tomography and also for training personnel. It is a light water reactor cooled by demineralised light water that flows through the reactor core by natural convection. The core is placed at the bottom of an open tank.In order to validate a computational thermal-hydraulic model of the coolant within the reactor tank, during normal operation of the TRIGA Mark II reactor, water temperatures were measured simultaneously at sixty different positions in the pool. The collected data allowed for a successful verification and validation of our model.In the paper, the CFD model setup and the associated numerical results are presented and validated against the experimental measurements of the temperature distribution within the pool.

305

Long-term system outage survey of the TRIGA reactor ViennaK. Mayer, Mario Villa, Helmuth BöckAtominstitutderÖsterreichischenUniversitäten,Stadionallee2,1020Wien,[email protected]

The TRIGA Mark II research reactor at the Vienna University of Technology – Atominstitut (VUT – ATI) became critical for the first time on March 7, 1962. Ever since this date precise logbooks were kept and all the operational data were recorded. These data also include component failures and system outages. Starting about five years ago all these outage information was evaluated in a systematic way, allocated to subsystems and analysed. To facilitate this procedure the main reactor systems were divided into 10 subsystems and each of these subsystems were further specified into smaller units. In this paper a survey of the outage events of the ten individual TRIGA subsystems (i.e. instrumentation and control system, ventilation system or primary cooling system) will be given together with it’s historical evolution. This also allows to carry out a system specific trend analysis and to optimize maintanance and replacement planning. Due to these analyses specific system replacements have been performed and will actually been carried out in the near future at the TRIGA reactor Vienna.

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306

JSI TRIGA 3D Reactor Model – Transition to Cartesian Geometry and Sample Kinetic Simulations

Vid Merljak1, Andrej Trkov2

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 InternationalAtomicEnergyAgency,Wagramerstr.5,P.O.Box100,A-1400Vienna,[email protected]

Recently, capabilities of performing kinetic simulations were implemented into the GNOMER deterministic code, which is a diffusion solver of the program package CORD-2. Various applications of the new capabilities are possible, among them the analysis of the rod insertion method for control rod worth determination. This method is routinely used at the JSI TRIGA research reactor and at the Krško NPP to measure control rod worth. For the purpose of code validation, a 3D model of the TRIGA reactor was developed. Special attention was required since the TRIGA’s cylindrical geometry had to be translated into a Cartesian geometry as required by the GNOMER code. The paper presents stages in the development of the model – namely, the cross sectional area preservation issues, the fuel element positioning, and the control rod macroscopic cross section generation – as well as results of few kinetic simulations and their comparison to the measured values. Calculated static and dynamic average radial power distributions are compared; static radial power distribution is also compared with the results of a dedicated 2D program package TRIGLAV and suggestions are made for further improvements of the model.

307

Analysis of the TRIGA Mark II Research Reactor Ex-Core Detector Response

Žiga Štancar1, Loic Barbot2, Luka Snoj1

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 CEA-CadaracheDEN/DER/SPEX/LDCI,Bâtiment238-Piece10,F13108Saint-Paul-lez-Durance,[email protected]

One of the most important operational safety parameters of the JSI TRIGA reactor is the reactor thermal power. The power measurements at the TRIGA research reactor are performed using ex-core instrumentation, ionization chambers contained in water-tight aluminum casings positioned at the outer edge of the graphite reflector surrounding the reactor core. There are five nuclear channels, each of them devoted to measurements of reactor power in a certain range. The widest power interval, i.e. from 100 mW to 300 kW, is covered by the linear channel, which is mostly used for thermal power monitoring at normal reactor operation conditions. The reactor reactivity and thus also the reactor power is regulated by insertion and withdrawal of four individually operated control rods. This can lead to situations where control rod insertion is asymmetric, which can significantly influence the ex-core detector (e.g. linear channel) response due to the so-called flux redistribution effect. Hence at the same nuclear power the readings on a single channel can vary by up to 20 % depending on the control rod configuration. Because the results of numerical reactor simulations are normalized to the reactor nuclear power it is of prime importance to take the flux redistribution effect into account and apply a correction factor to the ex-core power readings. The aim of the paper is to describe the correction factor which provides a theoretical estimate of the ex-core detector response deviance due to asymmetric control rod insertion together with an experiment performed to validate the latter factor. In the experiment a miniature fission chamber developed by CEA was inserted into the core of the TRIGA reactor and axial scans of absolute fission rates at three different control rod configurations were performed. The experiment was later modelled

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in detail with the Monte Carlo neutron transport code MCNP and a comparison between the measured and calculated absolute reaction rate profiles was made, thus directly verifying the accuracy of the correction factor. It was found that the factor does not adequately describe the effect when the control rods are either completely withdrawn or inserted, resulting in relative discrepancies between reaction rate measurements and calculations of up to 10 %. The observed disagreement served as a motivation for further computational analysis during which studies of the dependence of the flux redistribution effect and ex-core detector response on the configuration of the TRIGA control rods were made. Calculations of core flux profiles at over 100 control rod positions were performed, which enabled a better understanding of the flux redistribution factor and the application of the correction factor to the linear channel power readings.

309

Assessing Field Homogeneity: Application To Gamma Radiation Field Around Irradiated Nuclear Fuel

Junoš Lukan, Luka SnojJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Homogeneity of a field is a property often sought after in practical applications of various disciplines. In constructing a gamma irradiation facility by assembling spent nuclear fuel elements, for example, a homogeneous gamma field would be desired. There is no single definition of the concept, however, and one must be chosen when designing such a chamber.Existing, most widely used definitions of field homogeneity were reviewed. These include the ratio between the maximum and the mean or the minimum value and typical statistical moments, such as the median and the variance of the gradient of the field and the variance of the field itself. These measures were then tested with specific examples of fuel element arrangements. Additionally, a distribution of the largest differences between different field values and the accompanying percentiles were calculated. Since simulated field values are subject to modelling uncertainty, the propagation of error through calculation of these parameters was ascertained.A simplified model of an irradiation chamber built from TRIGA spent fuel elements was simulated in Monte Carlo neutron transport code MCNP MCNP and the gamma flux field was calculated for the purpose of benchmarking the measures of homogeneity. Different numbers of irradiated fuel elements were placed into water with an air-filled channel of cylindrical shape of varying diameter in the centre. The material specification of the fuel elements mimicked the one of a TRIGA fuel element, while the gamma-source distribution was obtained via the rigorous two-step method Chen2002,Davis2010 using FISPACT-II FISPACT.A linear combination of the homogeneity measures that explained the most variance of the studied examples was found by using principal component analysis. The simple ratio of the maximum and the mean value had the largest weight in this combination. Such a composite measure was also more robust to errors of the calculated field. Applicability of the measures to an irradiation chamber design was considered and practical solutions maximizing the gamma field homogeneity proposed.

MCNPJ. T. Goorley, M. R. James, T. E. Booth, F. B. Brown, J. S. Bull, L. J. Cox, … ,A. J. Zukaitis, Initial MCNP6 release overview: MCNP6 version 1.0 [LA-UR-13-22934],Los Alamos National Laboratory, 2013. Retrieved from \url{https://laws.lanl.gov/vhosts/mcnp.lanl.gov/pdf_files/la-ur-13-22934.pdf}

Chen2002Y. Chen, U. Fischer, ``Rigorous MCNP based shutdown dose rate calculations: Computationalscheme, verification calculations and application to ITER’’, Fusion Engineering and Design, 63–64, 2002,

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pp. 107–114. \doi{10.1016/S0920-3796(02)00144-8}

Davis2010A. Davis, R. Pampin, “Benchmarking the MCR2S system for high-resolution activationdose analysis in ITER”, Fusion Engineering and Design, 85(1), 2010, 87–92. 10.1016/j.fusengdes.2009.07.002

FISPACTJJ.-C. C. Sublet, J. W. Eastwood, J. G. Morgan, The FISPACT-II user manual:CCFE-R(11)11 Issue 6, UK Atomic Energy Authority, Culham Science Centre, Abingdon, 2014.Retrieved from \url{http://www.ccfe.ac.uk/assets/Documents/easy/CCFE-R(11)11.pdf}

310

Instrumentation and Control Implementations in Research Reactors: A Review

Andreas Ikonomopoulos, Melpomeni Varvayanni, Nicolas CatsarosNationalCenterforScientificResearch“DEMOKRITOS”InstituteofNuclearandRadiologicalSciencesandTechnology,EnergyandSafetyResearchReactorLaboratory,POBox60228,15310AgiaParaskevi,Attiki,[email protected]

The IAEA research reactor database identifies 160 operational or temporarily shutdown installations that achieved criticality more than forty years ago and 103 facilities that achieved first criticality 40 years ago or less. While a number of installations have been subjected to modifications and upgrades to satisfy the needs for higher neutron fluxes there are still facilities operating with parts and components of the original instrumentation and control (I&C) infrastructure that was installed during the initial construction. I&C is important to safety for being responsible for reactor control that involves startup, power regulation and shutdown as well as abnormal condition management. System obsolescence that is frequently accompanied by spare part unavailability may result in operational problems and extended reactor shutdown periods. In addition, novel requirements imposed by nuclear regulatory bodies and updated technical specifications may require the installation of additional instrumentation components along with the adoption of modern I&C implementations in research reactors. Significant technical advances in electronics and computer technology have occurred since the early I&C implementations in research reactors offering major functionality and performance improvements. It is noteworthy that these improvements come at significantly reduced costs making the combination of enhanced performance with low cost equipment suitable for assimilation by the research reactor community. The need for improved human-machine interfaces and on-line monitoring applications require a level of sophistication and reliability that can be met by the increased use of digital technologies. In that respect, IAEA has identified that: “From a technological viewpoint, developments in computers, such as in artificial intelligence, neural networks and display systems, have created opportunities for operator support that were not available in the past.” Formulating the blueprint of an instrumentation and control improvement in an ageing reactor installation requires careful consideration. To that end, an extensive review of state-of-the-art implementations proposed for research reactors is performed within the ARCADIA project (http://projectarcadia.eu/) with the intention to identify trends for I&C system applications in the field. The study will identify paradigms of computational intelligence implementations in research reactor facilities and study evaluate them with respect to enhanced accuracy, increased system safety, improved performance and extended availability.

http://nucleus.iaea.org/RRDB/RR/ReactorSearch.aspx?filter=0http://nucleus.iaea.org/RRDB/Content/Age/AgeHigh.aspxModern instrumentation and control for nuclear power plants: a guidebook. – Vienna: International Atomic Energy Agency, 1999

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Reactor Physics401

Simulation of the NPP Krško Core at Hot Full Power with CASL Core Simulator - VERA-CS

Fausto Franceschini1, Marjan Kromar2, Andrew T. Godfrey3

1 Westinghouse,RueMontoyer10,1000Bruxelles,Belgium2 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia3 OakRidgeNationalLaboratory,P.O.Box2008,OakRidge,Tennessee37831-6162,[email protected]

This paper describes the application of the Virtual Environment for Reactor Applications (VERA) core simulator (VERA-CS) under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL), to the core physics analysis of the Krško NPP. VERA-CS aims at enabling whole-core fuel cycle depletion deterministic transport analysis with subchannel thermal-hydraulic coupling. VERA-CS can also perform stochastic neutron transport calculations through a continuous-energy massively parallel Monte-Carlo code. This paper is focused on the application of VERA-CS to the analysis of the NPP Krško core at full power. COBRA-TF subchannel code is used for the calculation of the thermal-hydraulic parameters needed for the coupled calculation. Obtained pin-by-pin distributions are compared to the CORD-2 results.

402

Experimental study of the physical properties of ADS systems – measurement of high energy neutron fields by using the 89Y

threshold detectors Marcin Bielewicz, Elżbieta Strugalska-Gola, Stanisław Kilim, Marcin SzutaNationalCentreforNuclearResearch,ul.AndrzejaSołtana7,Otwock-Świerk,[email protected]

Study of deep subcritical electronuclear systems and radioactive waste transmutation using relativistic beams from the accelerator was performed. This work is a preliminary step toward the study of the physical properties of ADS systems (Accelerate Driven Systems). It is aimed to obtain fast neutron energy spectra inside the volume of the assembly using (n,xn) threshold reaction. Due to those experiments we can determine optimal energy for ADS systems and obtain the best position for transmutation processes. Normally in experiment like this one uses neutron activation foils made of gold (197Au), cobalt (59Co), bismuth (209Bi). New type foil, yttrium (89Y) was proposed. Yttrium detectors give good results for neutron energy from 10 MeV to about 50 MeV.Results of few experiments with 89Y foils are presented. The uranium experimental assembly “QUINTA” (consists of 512 kg of natural uranium rods arranged hexagonally and surrounded by a lead reflector) was irradiated with Dubna NUCLOTRON with 2, 4 and 8 GeV deuteron beam [1]. Neutron energy spectrum inside whole 3D model was obtained with 89Y threshold energy reaction. Also the comparison of the average neutron flux density per deuteron in three neutron energy ranges for the different deuteron beam is presented [2]. Cross section data for neutron induced reactions were calculated using the TALYS code. We have started checking experimentally those cross section values.

[1] W. Furman, M. Bielewicz, S. Kilim, E. Strugalska-Gola, M. Szuta, A. Wojciechowski et al. (2013) Recent

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results of the study of ADS with 500 kg natural uranium target assembly QUINTA irradiated by deuterons with energies from 1 to 8 GeV at JINR NUCLOTRON. Proceedings of Science, PoS (Baldin ISHEPP XXI) 086, 2013[2] M.Bielewicz, E. Strugalska-Gola, M. Szuta, A. Wojciechowski et al. Measurements of High Energy Neutron Spectrum (>10MeV) by Using Yttrium Threshold Foils in the U/Pb Assembly, Nuclear Data Sheets 119 (2014) 296–298

403

Sensitivity Analysis of Gas-cooled Fast ReactorJakub Lüley1, Stefan Cerba2, Branislav Vrban2, Ján Haščík1, Vladimir Nečas2

1 SlovakUniversityofTechnologyFacultyofElectricalEngineeringandInformationTechnology DepartmentofNuclearPhysicsandTechnology,Ilkovičova1,81219Bratislava,Slovakia2 SlovakUniversityofTechnology,FacultyofElectricalEngineeringandInformationTechnology, InstituteofNuclearandPhysicalEngineering,Ilkovičova3,81219Bratislava1,[email protected]

ALLEGRO, a 75 MWth reactor unit plays a vital role in the development of the electricity producing Gas-cooled Fast Reactor (GFR) prototype. As a demonstrator of the unique technology, never built before, it will serve to demonstrate the viability of the GFRs system. ALLEGRO design incorporates all the architecture and main materials and components foreseen for the GFR, except the power conversion systems. Within the reactor analysis and design calculation, sensitivity analysis offers a nuclear engineer a unique insight into the investigated system. Estimation of the change of the system response, due to change in some input parameter, can identify important processes and evaluate the influence of variation in this parameter. Sensitivity analysis of both configuration, ALLEGRO and GFR2400, was performed in order to identify main discrepancies between individual designs. Obtained results predict the way how to design and optimize the ALLEGRO core for testing of experimental sub-assemblies. ALLEGRO core is characterized by standard mixed oxide (MOX) subassemblies consisting of fuel pins with stainless steel cladding operating at an average coolant temperature around 400°C. In contrast, GFR2400 core is based on carbide pin fuel type with the application of refractory metallic liners used to enhance the fission product retention of the SiC cladding. Sensitivity analysis was performed using two computational tools to cover stochastic and deterministic approach of neutron transport calculation. In the first case, the TSUNAMI-3D code was utilized using ENDF/B-VII 238 group cross section data. Forward and adjoint transport calculations were carried out with KENO6 and the sensitivity coefficients were computed by the SAMS module. In the second case, self-developed perturbation PORK code was used which is interconnected with the diffusion flux solver DIF3D and ZZ-KAFAX-E70 based ENDF/BVII nuclear data library. To ensure correctness of the sensitivity coefficients, direct perturbation calculation for the most important nuclides was carried out.

404

Preliminary Analysis of The FLUOLE-2 ExperimentStephane Bourganel, Jacques Di-Salvo, Thiollay Nicolas, Soldevila MichelCEA,MemberofSNETPExecutiveCommittee,GifsurYvette91191,[email protected]

The FLUOLE-2 program is a benchmark-type experiment dedicated to neutron attenuation analysis with the aim of improving the TRIPOLI-4® Monte Carlo code validation. This two-year program has been developed by CEA (Commissariat a l’Energie Atomique et aux Energies Alternatives) to be representative of 900 and 1450 MWe Pressurized Water Reactors. For that purpose, different stainless steel structures

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have been designed and appropriately positioned inside the EOLE facility. EOLE is a pool type zero power reactor, composed of a cylindrical aluminum vessel with an over structure of stainless steel, able to contain various types of core and related structures. This critical mock-up is located at Cadarache CEA center. For the FLUOLE-2 experiment, the core has been designed as a 29×29 pins square lattice with both UOX and MOX fuel rods. Different kinds of dosimeters (cobalt, gold, tin, rhodium, indium, iron, nickel, titanium, aluminum, and vanadium) as well as different kinds of fission chambers (uranium 235, 238, and neptunium 237) have been irradiated inside and outside the core.The main objective of the FLUOLE-2 program is to validate the neutron attenuation through water and stainless steel structures. At mid-program, an appropriated modification of the core loading is done to ensure the irradiation of all the dosimeters by the two types of fuel rods (UOX and MOX). This allows an independent validation for the two main fission spectra inside the core: uranium 235, and plutonium 239.The purpose of this paper is to present a general overview of the FLUOLE-2 experiment as well as a preliminary analysis. The calculation scheme used to carry out this work is described. It is based on the TRIPOLI-4® 3D Monte Carlo code, and the DARWIN/PEPIN2 depletion code. TRIPOLI-4® is used to calculate the neutron source distribution in the core, the neutron propagation through the different structures, and fission rate values inside the fission chambers. The DARWIN/PEPIN2 code is dedicated to dosimeter activation calculations. All simulation tools used to carry out the FLUOLE-2 analysis presented in this article are developed by CEA, with financial support of EDF (Electricité De France) and AREVA. Nuclear data used by these codes are common, and based upon the JEFF3.1.1 and IRDF2002 cross section libraries to provide a uniform and consistent set of computational codes.The experimental database is provided by gamma scanning analysis as well as fission chamber and dosimeter activity measurements. Gamma scanning has been carried out on 176 fuel rods (UOX and MOX). It leads to the validation of the neutron source in the core. Fission chamber measurements allow the validation of the neutron source normalization. Dosimeter activity measurements provide information concerning the neutron attenuation in the different parts of the experimental device. Results concerning these three kinds of measurements are presented and discussed in this paper.

405

ALLEGRO uncertainty and similarity evaluation Branislav Vrban1, Stefan Cerba1, Jakub Lüley2, Ján Haščík2, Vladimir Nečas1

1 SlovakUniversityofTechnology,FacultyofElectricalEngineeringandInformationTechnology, InstituteofNuclearandPhysicalEngineering,Ilkovičova3,81219Bratislava1,Slovakia2 SlovakUniversityofTechnologyFacultyofElectricalEngineeringandInformationTechnology DepartmentofNuclearPhysicsandTechnology,Ilkovičova1,81219Bratislava,[email protected]

In the Gas Fast Reactor development plan, ALLEGRO is the first necessary step towards the electricity generating prototype GFR. This paper presents the first stage of study for the ESNII+ 75 MWth ALLEGRO reactor conceptual design with a brief description of the geometrical and material models used in the calculations.The main principles of fast reactor systems are rather well understood, however, their optimization, in order to comply more effectively with requirements and their timely deployment, requires the research in nuclear data. Although most nuclear data are by and large available in modern data files, their accuracy and validation is still a major concern. The main source of uncertainty in the calculated core responses is due to uncertainties in evaluated nuclear data such as microscopic cross sections (XS), fission spectra, neutron yield, and scattering distributions that are contained in cross section evaluations. These uncertainties are governed by probability distributions which are unknown, but the evaluated data values are assumed to represent the mean of the distribution, and the evaluated variance represents a measure of the distribution width. Correlations as well as uncertainties in nuclear data may have a significant impact on the overall uncertainty in the calculated response; thus, it is important to include them in the uncertainty analyses mainly in the new and never constructed reactor systems. TSUNAMI

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sequence available in SCALE system computes the contribution to the response uncertainty due to the cross-section covariance data with the use of system sensitivity profiles. The results of this uncertainty analysis of ALLEGRO ESNII+ MOX core are given and discussed in the paper. In addition the neutronic similarity of ALLEGRO MOX core to the several hundred critical benchmark experiments specified in the ICSBEP Handbook is evaluated by the use of three integral indices. According to the results, the similarity and uncertainty analysis of the ESNII+ ALLEGRO MOX core has identified specific problems and challenges in the field of neutronic calculations. The total uncertainty of calculated keff induced by XS data is in our calculation estimated to 1.04%. The additional margin from uncovered sensitivities was determined to be 0.28%The similarity assessment identified 9 partly comparable experiments where only one reaches acceptable values of similarity indices. Finally, it was demonstrated that TSUNAMI-IP utility can play a significant role in the future fast reactor development in Slovakia and in the Visegrad region.

406

Application of Monte Carlo Method for Burnup Dependent Full Core Neutronic Analysis of PBMR

Cihangir Çelik1, Mehmet Tombakoglu2

1 OakRidgeNationalLaboratory,P.O.Box2008,OakRidge,Tennessee37831-6162,USA-Tennessee2 HacettepeUniversity,NuclearEngineeringDepartment,06800Beytepe,Ankara,[email protected]

In this study, computational results of burnup dependent full core neutronic analysis of The Pebble Bed Modular Reactor (PBMR) are presented. The calculations performed by the MCNP-4b and BURN-HUNEM codes. BURN-HUNEM code was developed and utilized to perform depletion calculations using the cross section data generated by MCNP-4b to include double heterogeneity effect on neutronic parameters. The results for equilibrium cycle core composition and burnup presented for Once Through Then Out (OTTO) and Mehrfach-DUrchLauf (MEDUL) i.e., multi-pass , fuel cycle strategies.

407

Investigating a Newton-based, matrix-free, Neutronic-Monte Carlo/Thermal Hydraulic coupling scheme

Antonios Mylonakis1, Melpomeni Varvayanni2, Nicolas Catsaros2

1 NationalCentreforScientificResearch“Demokritos”,InstituteofNuclear&RadiologicalSciences& Technology,Energy&Safety,NuclearResearchReactorLaboratory,AgiaParaskeviAttikis, P.O.Box60037,15310Athens,Greece2 NationalCenterforScientificResearch“DEMOKRITOS”InstituteofNuclearandRadiologicalSciences andTechnology,EnergyandSafetyResearchReactorLaboratory,POBox60228,15310AgiaParaskevi, Attiki,[email protected]

In the field of nuclear reactor analysis, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance for both safety and design. So far in the context of Monte-Carlo neutronic analysis, the Gauss-Seidel iterative scheme, in a version for individual single-physics solvers, is mainly used for coupling with thermal-hydraulics. This work investigates the possibility of replacing the previous scheme with an approximate Newton algorithm. This method, called Approximate Block Newton, is actually a version of the Jacobian-free Newton Krylov method suitably modified for coupling mono-disciplinary solvers. Within this Newton scheme the linearized system is

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solved with a Krylov solver in order to avoid the need for creation of the Jacobian matrix. Main motivation for this approach is the interest for an algorithm that could maintain the distinct treatment of the involved fields within a tight coupling context. This work performs preliminary analysis in order to investigate the behavior of the proposed method in reactor analysis. More specifically, OpenMC, a Monte-Carlo neutronic code and COBRA-EN, a thermal-hydraulic code for sub-channel and core analysis, are merged in a coupling scheme using the Approximate Block Newton method aiming to examine the performance and the accuracy of this coupling scheme compared with those of the “traditional” sequential iterative scheme.

408

Evaluation of the Full Core VVER-440 Benchmarks Using the KARATE and MCNP Code Systems

György Hegyi1, Csaba Maráczy2, Gábor Hordósy2, Emese Temesvári2

1 Retired,Hungary2 CentreforEnergyResearchHungarianAcademyofSciences,P.O.Box49,H-1525Budapest,[email protected]

Evaluations have been performed for two calculational benchmarks described in the Atomic Energy Research (AER) association devoted to the VVER physics, using the models of the KARATE code system. The main task of these benchmarks is to test the assembly averaged and pin by pin power distribution predicted by codes that are used for reactor physical calculation in the engineering practice. The calculated power distribution especially in assemblies and pins next to the reflector strongly depends on the modelling choice. In case of a VVER-440 core high power peaking can be observed in the vicinity of control assembly, too. To study this phenomenon, a mathematical benchmark was defined by ŠKODA JS a.s. The proposal of this benchmark was presented at the 21st Symposium of AER in 2011. The reference solution has been calculated by the MCNP code using Monte Carlo method and the results have been published in the AER community. It is a 2D calculation benchmark based on the VVER-440 reactor core cold state. The core consists of fresh second generation Russian fuel assemblies with 3 different enrichments. One of the fuel assemblies has enrichment zoning and contains 6 gadolinium pins. There was no control assembly defined in the core. In case of the second exercise the working control group was fully inserted. The details about the geometry of the radial reflector and control assembly have been provided in the benchmarks.In the paper the two benchmarks will be outlined, the linear pin power calculation methodology of KARATE near the reflector and control assembly will be presented and our results will be shown.

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409

ANET Reaction Rates Validation Based on the VENUS-2 MOX Core Benchmark Analysis

Thalia Xenofontos1, Gregory Kyriakos Delipei1, Panayiota Savva1, Melpomeni Varvayanni2, Jacques Mailliard3, Nicolas Catsaros2, B. Gaveau4

1 NationalCentreforScientificResearch“Demokritos”,InstituteofNuclear&RadiologicalSciences& Technology,Energy&Safety,NuclearResearchReactorLaboratory,AgiaParaskeviAttikis, P.O.Box60037,15310Athens,Greece2 NationalCenterforScientificResearch“DEMOKRITOS”InstituteofNuclearandRadiologicalSciences andTechnology,EnergyandSafetyResearchReactorLaboratory,POBox60228,15310AgiaParaskevi, Attiki,Greece3 InstitutduDéveloppementetdesRessourcesenInformatiqueScientifiqueCNRS,Orsay,France4 UniversitéPierreetMarieCurie,75005Paris,[email protected]

ANET is a stochastic neutronics code under development aiming to simulate GEN II/III reactors, as well as innovative nuclear reactor designs such as Accelerator Driven Systems. ANET’s reliability in performing criticality and fluence rates computations has been successfully tested in previous works. The OECD/NEA VENUS-2 MOX international benchmark is used here in order to proceed with the ANET code benchmarking and validation as far as the reaction rates simulation is concerned. The VENUS facility is a zero power critical reactor which core comprises three types of fuel pins, i.e. 3.0wt% and 4.0wt% enriched in 235U as well as mixed oxide 2.0wt% and 2.7wt% enriched in 235U and high-grade plutonium respectively. In the framework of the VENUS-2 MOX core benchmark fission rate measurements at 21 different vertical planes were carried out in six fuel pins, two of each fuel type, in representative core positions. ANET has performed fission rate simulations in all suggested positions for both 235U and 238U isotopes following the axial measurements protocol. The obtained results are compared with experimental data and computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI and MCNP and are found in satisfactory agreement with both, verifying thus ANET’s ability to successfully simulate important parameters of critical systems.

410

Determination of the NPP Krško Spent Fuel Activity Marjan Kromar1, Bojan Kurinčič2

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 NuklearnaelektrarnaKrško,Vrbina12,8270Krško,[email protected]

Nuclear fuel is after it's use in the reactor stored in the pool. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the inventory is important during the storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel activity with the Serpent code is investigated. Serpent is a well known Monte Carlo code used primarily for the calculation of the neutron transport in the reactor. It has been validated for the burn-up calculations. In the calculation of the fuel activity different set of isotopes is important than in the neutron transport case. A few typical cases of the NPP Krško fuel are selected for comparison with the Fispact code. Fispact is a well known inventory code capable of performing modeling of activation, transmutations and burn-up induced by neutron, proton, alpha, deuteron or gamma particles incident on matter. However, it is a 0-dimensional code and can not capture

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directly spatial effects. Comparison with the Serpent code is performed to verify that the Serpent is taking into account all isotopes important to asses the fuel activity. After the code validation a sensitivity study is carried out. Fuel isotopic composition is namely pretty dependent on the neutron spectrum. Influence of several factors such as boron concentration, burnable poison presence, water density etc. is analyzed.

411

Severe accident gamma dose distribution through NPP Krško containment and Auxiliary Building calculated using SCALE6.1/

MAVRIC sequenceMario Matijević, Davor Grgić, Dubravko PevecUniversityofZagreb,FacultyofElectricalEngineeringandComputing,Unska3,10000Zagreb,[email protected]

The ORNL SCALE6.1 code package was used for the Monte Carlo model preparation of the NPP Krško corresponding to reactor, primary loop components, containment and adjoining buildings. A fairly detailed model of the reactor with concrete compartments housing primary pumps and steam generators was developed using as-built dimensions. Auxiliary buildings were modeled as empty at the moment, taking into account the bulk dimensions of concrete walls and floors. Analyses of gamma dose distribution in the case of hypothetical SBO accident was performed using hybrid deterministic-stochastic FW-CADIS methodology of the MAVRIC shielding sequence. Preparation of the gamma source was done using isotopic concentrations calculated with RADTRAD code and 18-groups ORIGEN energy structure. The source is homogenous and distributed over all air regions in a containment. The influence of melted material in reactor cavity is not taken into account. Under such conditions the aim was to calculate gamma dose rates through different floors, sections and buildings, posing significant shielding problem. Special attention was given to gamma doses in the containment and in AB rooms close to the containment. For that purpose, the integrated discrete ordinates code Denovo, based on the adjoint transport theory, was used for calculation of variance reduction parameters (weight windows and biased source) over the XYZ meshes covering the problem domain. Denovo utilizes Koch-Baker-Alcouffe parallel transport sweep algorithm and Krylov iteration on multigroup equations giving space-energy dependent fluxes and variance reduction parameters. These variance reduction parameters, that work in tandem, are automatically transferred to the functional module Monaco for the final, optimized Monte Carlo transport. Different adjoint source locations were investigated, giving optimized dose rates over localized or distributed phase-space regions. Visualization of the obtained results in 3D was done using VisIt code from the LLNL.

412

Use of LiF-TLD100 Detector with B4C filter in Neutron Dosimetry

Ilkem Aydogan, Ayhan YilmazerHacettepeUniversity,NuclearEngineeringDepartment,06800Beytepe,Ankara,[email protected]

In this study, use of LiF-TLD100 as a detector material with B4C filter in neutron personnel dosimetry is investigated both experimentally and by Monte Carlo simulation. In the first part of the study, cylindrical thermoluminescence LiF-TLD100 crystals are irradiated by an Am-Be neutron source. Then, the absorbed radiation dose in the crystals is determined in TLD reader. Similar measurements are made by locating the B4C filters, which are made sintering boron carbide that includes natural rate of B-10, in front of

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LiF-TLD100 crystals. The absorbed radiation doses after filtering are again determined in TLD reader.The absorbed radiation doses in the crystals are also calculated by using Monte Carlo simulations done by MCNP code. Consistent absorbed dose values between experimental and calculated results are observed. It is demonstrated that B4C is a proper material that could be used as a filter in personal neutron dosimetry. Moreover, the limits of sensitivity to neutron particle of LiF-TLD100 detectors is observed to be satisfied in order to use them in neutron dosimetry.

413

Simulation of fuel cycle for Krško NPP using Monte Carlo code and GNOMER diffusion code

Dušan Ćalić1, Andrej Trkov2

1 ARJE,analizeinraziskavenapodročjujedrskeenergetike,d.o.o.,Vrbina17,8270,Krško,Slovenia2 InternationalAtomicEnergyAgency,Wagramerstr.5,P.O.Box100,A-1400Vienna,[email protected]

This paper present the use of Monte Carlo code Serpent 2 as a cell calculational tool for producing the cross sections that will be used to perform full-core calculations using GNOMER code. In the past the effective homogenization technique using Monte Carlo Serpent code was already demonstrated with the conclusion that this procedure could be used and replace the current deterministic technique to obtain cell averaged cross sections [1]. We have developed and implemented new procedures together with the burnup calculations by using new predefined ISOlib library, based on Monte Carlo burnup calculations to perform and predict the Krško NPP key core parameters, by analysing the comparison vs. measurements and current CORD-2 calculations. The results for fuel Cycle 1 will be presented in this paper.

[1] Ćalić D., Kromar M, Trkov A., Neutron Multigroup Homogenized Cross Section Determination with the Monte Carlo Method, Nuclear Energy for New Europe 2012, Ljubljana, Slovenia, 2012.

414

Application of Support Vector Regression Method on Neutron Buildup Factors

Paulina Dučkić, Krešimir Trontl, Dubravko PevecUniversityofZagreb,FacultyofElectricalEngineeringandComputing,Unska3,10000Zagreb,[email protected]

In this paper Support Vector Regression (SVR) method application on neutron buildup factors determination is investigated. When applied on gamma-rays, SVR method shows very good results, leading to the conclusion that there is a potential of using this approach for practical gamma-ray buildup factors determination within established point kernel codes, like QAD-CGGP. A possibility of SVR application on neutron buildup factors estimation has yet to be determined. Neutrons can interact with shielding materials in many ways, including scattering and absorption processes, as well as production of secondary neutrons, protons, gamma-rays or alpha particles. The mentioned by-products can further interact with shielding materials and therefore significantly increase the measured quantity. Additionally produced neutrons can be described with neutron buildup factors. For this study, neutron buildup factors are calculated using approximate formula and are strongly dependent on the shielding material, thickness, and the incident neutron energy range. Typical shielding materials like iron and concrete are observed, for neutron energy range 0.5 eV-14 MeV, and 1-10 mfp thicknesses. Hence, the input vector is multi-dimensional, comprised of the atomic number of the material, shielding thickness, and the incident neutron energy.

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Output vector is one-dimensional, presenting target buildup factors. Due to high complexity of neutron transport through shielding material, building the SVR model for neutron buildup factor determination is very demanding. When regarding the size of the training set, one is confronted with a vast number of potential training points. In order to minimize the number of training points and speed up the training process, active learning measures are applied. By combining various active learning methods, the training set is composed of most informative points, leading to good generalization properties of the model with adequate accuracy.

415

Investigation of the Allegro MOX Pin Core design by stochastic and deterministic methods

Stefan Cerba1, Branislav Vrban1, Jakub Lüley2, Jan Jascik1, Vladimir Nečas1

1 SlovakUniversityofTechnology,FacultyofElectricalEngineeringandInformationTechnology, InstituteofNuclearandPhysicalEngineering,Ilkovičova3,81219Bratislava1,Slovakia2 SlovakUniversityofTechnologyFacultyofElectricalEngineeringandInformationTechnology DepartmentofNuclearPhysicsandTechnology,Ilkovičova1,81219Bratislava,[email protected]

The Gas-cooled Fast Reactor (hereinafter the GFR) is one of the six most promising reactor concepts selected by the Generation IV International Forum (GIF). The design of this reactor may partially benefit from a number of previously proposed but not realized conceptions as well as from the research of related technologies of the Sodium-cooled Fast Reactor (SFR) and the Very High Temperature Reactor (VHTR). There have been numbers of research projects done into the GFR technology, however no one has led to the final built of a gas cooled fast reactor so far, therefore the research of the ALLEGRO GFR demonstrator is one of the most necessary steps in the development of the GEN IV GFR. ALLEGRO is small helium cooled 75 MWth unit, without an option to produce electricity. The main objective is to demonstrate the key GFR technologies and to perform tests of innovative fuel and structure materials. Although the active ore of a large GFR 2400 unit is assumed to be composed of ceramic materials, but in the first ALLEGRO core MOX fuel will be used, to demonstrate the viability of technology at low temperatures. In the next step also ceramic core materials will be investigated. The presented analysis deals with the investigation of the MOX pin core design of the ALLEGRO reactor. To ensure diversity of results, the calculation scheme includes codes using both stochastic and deterministic approaches. As a reference stochastic code the SCALE6 system with 238 group ENDF/B VII.0 XS data has been selected. For the justification of these stochastic results the deterministic diffusion DIF3D code seems to be a suitable solution. For the DIF3D calculations 2 sets of MG XS libraries were prepared. In the first case a new set of 620 group XS was created based on ENDF/B VII.1 evaluated data and the average neutron spectrum from the GFR 2400 core was used as weighting function. In the second case the Korean ZZ_KAFAX_E70 XS library was used. In the analyses also collapsed group structures were used. The main investigated parameters include excess reactivity, neutron spectra studies, control rod worth and interference as well as local multiplication factors (LMF).

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416

Reactivity power and temperature coefficients determination of the TRR

Ahmad LashkariAtomicEnergyOrganizationofIran,KHAfrica,K.Tandis,No.7P.O.Box14155-1339,19156Tehran,[email protected]

The aim of this paper is present the experimental results of the power and temperature coefficient of reactivity of the Tehran Research Reactor (TRR) at the Nuclear Science and Technology Research Institute (NSTRI) of Iran. In this work in addition to the previous method, new methods were used to measure the reactivity coefficient of TRR. The experiments were performed in the TRR reactor with 33 MTR fuel elements in the core. At the first method, we determined the isothermal coefficient of TRR and then calculated power and temperature coefficient of reactivity. This method is very similar to the method that used to determine the power coefficient of IPR-R1 TRIGA reactor. One of the new methods used in this study is comparing the situation of control rod positions in two cooling modes The aim of this paper is present the experimental results of the power and temperature coefficient of reactivity of the Tehran Research Reactor (TRR) at the Nuclear Science and Technology Research Institute (NSTRI) of Iran. In this work in addition to the previous method, new methods were used to measure the reactivity coefficient of TRR. The experiments were performed in the TRR reactor with 33 MTR fuel elements in the core. At the first method, we determined the isothermal coefficient of TRR and then calculated power and temperature coefficient of reactivity. This method is very similar to the method that used to determine the power coefficient of IPR-R1 TRIGA reactor. One of the new methods used in this study is comparing the situation of control rod positions in two cooling modes (natural and force) in the same power of TRR. The difference between two control rod configurations is caused by the temperature difference in coolant in two modes. With measuring the difference reactivity and coolant temperature, we can calculate reactivity coefficient.The last new method that is much more efficient than the above methods, using the dynamic behavior of reactor power due to change of reactor core temperature. The main advantage of this method is that we can measure the reactivity coefficient of reactor very fast and independent of the control rods worth and positions. The average values of the temperature and power reactivity coefficient of the fuel and the coolant in TRR are: α_T (F)=1.95 pcm/0C, α_T (m)=13.57 pcm/0Cα_p (F)=0.16 pcm/KW, α_p (m)=0.89 pcm/KW

417

The New Edition of Karlsruhe Nuclide Chart in Summer 2015Zsolt Soti1, Magill Joseph2, Dreher Raymond2, Pfennig Gerda2

1 InstituteforTransuraniumElements,P.O.Box2340,D-76125Karlsruhe,Germany2 NucleonicaGmbHc/oEuropeanCommission,Hermann-von-Helmholtz-Platz1,76344Eggenstein- Leopoldshafen,[email protected]

For almost 60 years, the Karlsruhe Nuclide Chart has provided scientists and students with structured, accurate information on the half-lives and decay modes of radionuclides, as well as energies of emitted radiation. This tradition continues with the release of the new 9th Edition in 2015. Since the 2012 edition, more than 100 nuclides have been discovered and about 1400 nuclides have been updated. In summary, the new edition contains decay and radiation data on approximately 3230 ground state nuclides and 740

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isomers from 118 chemical elements. The accompanying booklet provides a detailed explanation of the nuclide box structure used in the Chart. An expanded section contains many additional nuclide decay schemes to aid the user to interpret the highly condensed information in the nuclide boxes. The paper-based Karlsruhe Nuclide Chart, with its fold-out, wall chart and auditorium chart versions, remains an aesthetically appealing record of human achievement in nuclear science. It provides a unique overview of current knowledge in nuclear field. The new online version - The Karlsruhe Nuclide Chart Online - was published in 2014 contains a number of additional features such as the facility to compare nuclide data from different editions, etc.

418

Scale 6.1.3 evaluation of the heavy reflector effective cross sections of a GEN III PWR system and a Serpent 2.1.23 model

comparisonAntonio Guglielmelli1, Federico Rocchi2, Marco Sumini3

1 NuclearEngineeringLaboratoryofMontecuccolinoUniversityofBologna,viadeiColli16, 40136Bologna,Italy2 ENEA,ViaMartiridiMonteSole4,40129Bologna,Italy3 UniversityofBologna,FacultadiIngehneria,VialeRisorgimeno2,40136Bologna,[email protected]

A rigorous calculation of GEN III PWR effective cross sections for a heavy reflector, by means of a proper geometry and material modelization, is a fundamental task for a correct evaluation of the radial power distribution in the whole core. This need is due to the circumstance that in a nuclear core equipped with a massive stainless steel reflector there is a strong modification of the radial power distribution with respect to the same core equipped with a standard reflector whose precise assessment depends on a rigorous evaluation of the reflector zone cell data. For this reasons, the paper proposes the calculation results of ADFs and homogenized two group condensed effective heavy reflector cross sections in the central and peripheral zones of a GEN III PWR core. The effective cross sections evaluation has been performed by means of the T-NEWT control module of SCALE 6.1.3 modular code system. The detailed calculation procedure has been based on a rigorous procedure using the functional sequence CENTRM/PMC for self-shielding calculation and on the NEWT functional module for transport calculation. The cross section library used is the v7-238 based on the ENDF/B-VII (Release 0) with 238 energy groups (148 fast and 90 thermal). Specifically, it has been realized a preliminary study to verify the different quantitative neutronic behavior – in terms of backscattering neutron current at the assembly/reflector interface – between a heavy and a standard reflector. Afterwards, a series of calculations have been carried out to verify the influence of geometric and material assembly/reflector models (homogeneous, homogenized slabs and heterogeneous) on the effective cross sections numerical results. Subsequently, it has been investigated the consequence of the use of 2 or 8.5 assemblies coupled with the reflector zone on the effective cross sections values. The last part of the paper shows the cell data results for central and peripheral areas of the GEN III heavy reflector. These calculations have been performed using the geometric and material parameters taken from U.S. EPRTM data. Finally, it also has been carried out an analysis to investigate the 2-D effect in the angular reflector region on the effective cross sections results by means of 2x2 color set modelization. The outcomes of the T-NEWT reference case are at last compared with those obtained with the Serpent 2.0 code.

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419

Scale 6.1.3 effective heavy reflector cross sections sensitivity analysis for a PWR GENIII assembly/reflector system

Antonio Guglielmelli1, Federico Rocchi2, Marco Sumini3

1 NuclearEngineeringLaboratoryofMontecuccolinoUniversityofBologna,viadeiColli16, 40136Bologna,Italy2 ENEA,ViaMartiridiMonteSole4,40129Bologna,Italy3 UniversityofBologna,FacultadiIngehneria,VialeRisorgimeno2,40136Bologna,[email protected]

The neutronic simulation results – obtained by means of a nodal diffusion code – of a PWR GENIII core equipped with a heavy reflector, strongly depend on the choice of an “adequate” set of effective reflector cross sections parameters previously determined with a cell transport code. The “adequacy” regards, first of all, the use of a proper sets of code parameters to perform a rigorous heavy reflector cross sections calculation and then the possibility of choosing a set of reflector cell data that match the same operative conditions of the core.For this reason, the paper presents the results of a sensitivity analysis on code and operative parameters for an assembly/reflector PWR GENIII system. The purpose of this work has been the production of a set of problem-dependent effective heavy reflector cross sections for the central zone of a PWR GENIII system that may be used for a nodal diffusion code calculation. The deterministic codes used for the calculation are CENTRM (1-D transport calculation module using ENDF-based point data on an ultrafine grid composed of about 3000-7000 energy points); PMC (module to generate problem-dependent self-shielding cross sections) and NEWT (module for 2-D deterministic transport calculation). Each of the above codes have been managed by the T-NEWT control module of the SCALE 6.1.3 package. The cross sections libraries used are the v7-238 based on the ENDF/B-VII (Release 0) with 238 energy groups (148 fast and 90 thermal). In detail, the paper presents the results of a sensitivity analysis on the effective reflector cross sections for a series of code parameters such as: reflector zone computational meshes, Sn and Pn transport parameter, cross sections libraries and convergence parameters (eigenvalue and eigenfunction). It is also presented the result of a sensitivity analysis on the operative parameters such as boron concentration (0-1300 ppm), operative conditions (HZP and HFP) and reflector temperature (Tmax and Tmod).

420

BEACON Version 6 vs 7 Software Code Comparison in NPP Krško Operating Cycle 27

Matjaž Božič, Martin Chambers, Bojan KurinčičNuklearnaelektrarnaKrško,Vrbina12,8270Krško,[email protected]

The BEACON (Best Estimate Analysis of Core Operations – Nuclear), is a core monitoring and operational support package developed by Westinghouse for use at PWR plants. The BEACON system is a real-time core monitoring system which uses existing core instrumentation data and an on-line nodal neutronics model to provide continuous core power distribution monitoring. The monitoring function uses plant instrumentation to develop and provide information on the actual core conditions. Accurate core prediction capabilities are also available through BEACON. These prediction functions utilise a 3D nodal model that is continuously updated to reflect the plant operating history. The BEACON functions allow the access to the continuously updated nodal model to perform automated calculations of estimated critical

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conditions for start-up, shutdown margin calculations for maintaining the minimum boron concentration after reactor trip, core depletion with optional coast down calculations, and load manoeuvre predictions.Currently, NEK is transitioning from BEACON TSM version 6 to 7, during NEK operating cycles 27-28 both versions are operational and a comparison or benchmarking between the code performance was made specific to NEK. Here, the performance in key areas of interest to NEK are examined and discussed.

421

Participation in OECD/NEA Oskarshamn-2 (O2) BWR Stability Benchmark for Uncertainty Analysis in Modelling Using Triton and Keno for Transport Calculations and Tsunami and Sampler

for Cross Section Error PropagationAntonella Labarile1, Teresa Barrachina2, Rafael Miró3, Gumersindo Verdú4

1 InstituteforIndustrial,RadiophysicalandEnvironmentalSafety(ISIRYM)5KBuild-Universidad PolitecnicadeValencia,CaminodeVera,s/n,46022Valencia,Spain2 DepartmentofChemicalandNuclearEngineering,PolytechnicUniversityofValencia,CamídeVerasn, 46022Valencia,Spain3 UniversitatPolitecnicadeCatalunya,C.JordiGirona,31,08034Barcelona,Spain4 UniversidadPolitecnicadeValencia,DepartamentodeIngenieríaQuímicayNuclear,Caminode Veras/n,46022Valencia,[email protected]

Nowadays, with an increased number of light water reactor (LWRs) around the world, there is a largest interest in improving safety analysis research with Best-Estimate computer code, to model the complex system of nuclear power plants for making predictions that includes quantitative uncertainty analysis.This work is related with OECD/NEA coupled code benchmarks based on operating reactor data, which objective is to establish confidence bound calculation as a usual practice in extending code applications, from its original intended use to more challenging events like unstable power oscillations without scram.Benchmark references are based on measured plant data of boiling water reaction (BWR) Oskarshamn-2 Nuclear Power Plant that experienced a stability event on February 25, 1999. However, a Steady-state calculations in operating condition was performed in this work with aim to obtain the keff and cross section values as well as sensitivity analysis and uncertainties results. Calculations were carried out for BWR fuel elements, in two different configurations (with and without control rod) at Hot Zero Power (HZP) state.Neutronics calculations were accomplished with computation of energy collapsed and homogenized macroscopic cross sections through SCALE-6.2beta3 program. The deterministic lattices modelling are carried out using TRITON/NEWT for transport calculations, while a Monte Carlo approach is carried out with TRITON/KENO module with purpose to compare keff results in deterministic and Monte Carlo calculation.The uncertainties and sensitivity analysis of cross sections calculation has been performed using the TSUNAMI module, which uses the Generalized Perturbation Theory (GPT) and SAMPLER, which makes use of stochastic sampling techniques of cross sections perturbations. Perturbed values for flux in each energy group are applied and finally the sensitivity coefficients are calculated to discriminate the most sensitive and influential parameters in results of the keff and macroscopic and microscopic cross sections. The results of the uncertainty analysis will be part of the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) to perform uncertainty analysis of the BWR stability prediction.

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422

TRITON vs POLARIS. Comparison between two modules for LWRs modelling in SCALE6.2

Antonella Labarile1, Teresa Barrachina2, Rafael Miró3, Gumersindo Verdú4

1 InstituteforIndustrial,RadiophysicalandEnvironmentalSafety(ISIRYM)5KBuild-Universidad PolitecnicadeValencia,CaminodeVera,s/n,46022Valencia,Spain2 DepartmentofChemicalandNuclearEngineering,PolytechnicUniversityofValencia,CamídeVerasn, 46022Valencia,Spain3 UniversitatPolitecnicadeCatalunya,C.JordiGirona,31,08034Barcelona,Spain4 UniversidadPolitecnicadeValencia,DepartamentodeIngenieríaQuímicayNuclear,Caminode Veras/n,46022Valencia,[email protected]

One of the most important challenges in reactors research is the development of best estimate codes that enable to reduce uncertainties in calculations and increasing results reliability. Moreover, S&U analysis and code validation are needed. This paper provides a comparison of two modules of SCALE in a Pressurized Water Reactor simulation.Based on measured data from Three Mile Island-1 reactor it has been calculated the Keff and the cross sections with two modules in SCALE: TRITON and POLARIS. TRITON is a validated code for transport calculation while POLARIS is in distribution since the end of 2014. The aim is to compare the results of the Keff and cross sections in two simulations, and perform sensitivity and uncertainty analysis in obtained results.The TRITON and POLARIS calculations were performed for fuel elements of PWR Three Mile Island-1, in two different configurations (with and without control rod), in Hot Zero Power (HZP) and Hot Full Power (HFP) condition.The results are presented in this paper. It was found a good correlation between the results of the two simulations and, in addition, POLARIS module presented lower computational times and good stability in its parameters. After having compared the obtained results, a sensitivity analysis has been conducted to confirm the validation of the two modules and to study the influence of the uncertainties in the fuel element calculation.

423

Neutron noise analysis in the NPP Krško - Comparison of Cycles 26, 27 and 28

Marjan Kromar1, Bojan Kurinčič2, Urban Simončič1, Rok Bizjak2

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 NuklearnaelektrarnaKrško,Vrbina12,8270Krško,[email protected]

NPP Krško has a few years ago initiated the neutron noise monitoring program in which some measurements of the nuclear instrumentation signals (in-core and ex-core detectors signals) are executed during the operational cycle. With the help of noise diagnostic methods, the vibrational frequencies of the reactor internals and fuel components as well as the axial distribution of the neutron noise are determined. Comparison of characteristic vibrational frequencies over several cycles can detect vibrational changes indicating possible unwanted reactor behavior. In cycle 28 NPP Krško has performed an Upflow Conversion (UFC) modification on the reactor vessel assembly in order to reduce the potential for baffle jetting. The flow direction in the baffle-barrel region has been redirected from downflow to upflow regime. In

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this paper evaluation of the measurement data is presented, which was performed to determine any significant changes in vibrational frequencies and to help in the ongoing development process to establish more efficient measurement procedure.

424

An Overview of the Improvements in Fuel Cycle Sustainability for GEN III+ Reactors

Melpomeni Varvayanni, Andreas Ikonomopoulos, Nicolas CatsarosNationalCenterforScientificResearch“DEMOKRITOS”InstituteofNuclearandRadiologicalSciencesandTechnology,EnergyandSafetyResearchReactorLaboratory,POBox60228,15310AgiaParaskevi,Attiki,[email protected]

In the upcoming decades huge quantities of energy will be needed worldwide, especially in the form of electricity generated in an environmentally friendly manner. Nuclear power is the most environmentally propitious way of producing electricity at a large scale ensuring, at the same time, stability of supply and considerable reduction of the carbon dioxide levels in the atmosphere. Moreover, nuclear energy has a the potential of playing a significant role in a sustainable global energy supply by developing technical and institutional innovations within the context of a sustainable nuclear fuel cycle. In this respect, the thorium fuel cycle appears to offer long-term energy security benefits due to its potential for being a non-proliferative, self-sustaining fuel that can be used in various types of nuclear reactors (both operational in operation and under design) suggesting, therefore, an important and potentially viable technology that seems able to contribute to the formation of credible, long-term, nuclear energy scenarios. Towards nuclear sustainability, reactor designers may affect the nuclear fuel cycle by building reactors that simply produce electricity or combine electricity production with burning a few of the fission process by-products. In fact, it is foreseen that within the next few decades the continued deployment of Generation III/III+ reactors will be accompanied by with the phasing out of all but the newer Generation II designs. Furthermore, light water reactors (LWRs) have remained the predominant reactor type worldwide while they are expected to continue dominating up to the latter part of the century. The present work performed within the ARCADIA project (http://projectarcadia.eu/) provides an overview of the improvements proposed for fuel cycle sustainability for GEN III+ LWRs attempting also to localize computational needs while focusing, mainly, towards two directions: (a) the utilization of MOX fuel containing thorium and (b) the achievement of higher-than-conventional conversion ratios through the concept of reduced-moderation cores. The definition adopted for the term “sustainability” throughout the current review is based on the “IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles” and refers to certain key elements, such as environment, resource utilisation, waste management, infrastructure, proliferation resistance and physical protection, safety and economics.

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425

Ringhals-1 BWR stability analysis with TRACE/PARCSConsuelo Gómez-Zarzuela Quel1, Agustin Abarca2, Teresa Barrachina3, Rafa Miró4, Gumersindo Verdú2

1 UniveristatPolitecnicadeValencia,CamídeVeras/n,46022Valencia,Spain2 UniversidadPolitecnicadeValencia,DepartamentodeIngenieríaQuímicayNuclear,Caminode Veras/n,46022Valencia,Spain3 UniversitatPolitecnicadeCatalunya,C.JordiGirona,31,08034Barcelona,Spain4 DepartmentofChemicalandNuclearEngineering,PolytechnicUniversityofValencia,CamídeVerasn, 46022Valencia,[email protected]

This work is framed in nuclear safety field in BWR. This kind of nuclear power plants has suffered several instability cases, whose dominant mode is the one that combines neutronics and thermal hydraulic effects. In order to realize different studies about them, the analyst requires computer codes that have to be able to represent with high reliability the physical phenomena that take place during an instability. In addition, computer codes are necessary to the reactor licensing, following the SRP guide from the NRC. Previously, these computer codes have to be validated. The aim of the present work is to collaborate in the validation of the methodology TRACE v5.0/ PARCS v3.2 for instability simulations. In particular, we reproduced the results obtained in the NEA Ringhals I BWR Stability Benchmark and compared to those obtained for the core only and the results obtained for the complete vessel. The operational point selected is Record 9 from cycle 14. The core model was modeled with 72 thermal-hydraulic channels, 71 represent the active core and one representing the core bypass. In order to design this configuration, the azimuthal modes have been taken into account. With this mapping, the possibility of conditioning the oscillation configuration is avoided. The vessel model includes also pumps, downcomer, separator and dryers. The perturbation carried out consists of a control rods movement, which causes a larger excitation to the subcritical modes, leading to the expected out-of-phase oscillation. Therefore, the results allow the validation of the procedure TRACE v5.0/PARCS v3.2 for stability analysis.

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Severe Accidents & Probabilistic Safety Assessment



501

The Latest Results from Source Term Research: Overview and Outlook

Luis E. Herranz1, Tim Haste2, Teemu Kärkelä3

1 CIEMAT,Avda.Complutense,22,28040Madrid,Spain2 InstitutdeRadioprotectionetdeSuretéNucléaire,Bât.702CentredeCadarache,BP3-13115SaintPaul lezDurance,France3 VTTTechnicalResearchCentreofFinland,Tietotie3,Espoo,02044VTT,[email protected]

Source term research has continued internationally for more than 30 years, with the overall aim of increasing confidence in the methods used in calculating the potential radioactive release to the environment after a severe reactor accident. Important data have been obtained from small- and large-scale experiments, mainly under international frameworks such as OECD/NEA, European Framework Programs of EURATOM and specific consortia. In particular, Phébus FP and associated studies provide outstanding insights into fission product release and transport and, particularly, containment iodine chemistry, which have been and are being encapsulated in recent versions of severe accident analysis codes like ASTEC 2.1, MELCOR 2.1 and MAAP-EDF, while data from newer projects such as VERDON, BIP and THAI are being interpreted with a view to further improvements in code capability.This paper briefly synthesizes the recent main outcomes from source term research concerning the above topics, and also source term mitigation. It highlights the knowledge gaps remaining and discusses ways to proceed, addressing those items considered high priority, taking as a basis the most recent source term workshop held in April 2015 under the international non-profit association for Gen. II and Gen. III reactor research, NUGENIA, sub-task 2.4 (Source Term) of technical area 2 (Severe Accidents). Two major issues generally affect experimental data from source term research: analytical exploitation and scale-up. Concerning fission product release, oxidizing environments potentially leading to increased release of harmful nuclides, like Ru-103/106, are considered of utmost interest. For transport, potential revolatilisation of fission products embedded in deposits needs further study. Containment iodine chemistry is and has been extensively investigated in recently ended, ongoing and upcoming OECD projects, like BIP3, STEM2 and THAI3, and the net results need evaluation. For mitigation, long term filter behavior regarding fission products, existing capabilities to remove Ru and the scrubbing capacity of pools undergoing saturation need further research. Aside from further knowledge-driven research, there is consensus on the need to assess the source term predictive ability of current system codes.

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502

In Vessel Corium Propagation Sensitivity Study Of Reactor Pressure Vessel Rupture Time With PROCOR Platform

Maciej Skrzypek1, Eleonora Klara Skrzypek1, Laurent Saas2, Romain Le Tellier2

1 NationalCentreforNuclearResearch,ul.AndrzejaSołtana7,Otwock-Świerk,Poland2 CEA,MemberofSNETPExecutiveCommittee,GifsurYvette91191,[email protected]

The problem of the corium propagation for PWRs in the Reactor Pressure Vessel and the time of the RPV failure is one of main issues of study in area of severe accidents. The PROCOR numerical platform created by the CEA severe accident laboratory is modeling corium propagation for LWRs, its relocation to the Lower Plenum and RPV failure. The idea behind the platform was to provide the tool that will be sufficiently fast to be able to perform numerous calculations in reasonable time frame. Therefore the work on the development of the models, describing in-vessel issues, is continuously performed through the simplified phenomena modeling, their verification and sensitivity studies. The recent activities, in scope of PROCOR development, involved cooperation between French CEA experts and Polish PhD students, who were engaged in the topics of core support plate modeling and analysis of the phenomena of thin metallic layer on the top of the corium pool. Those issues were identified to strongly influence on the course of the severe accident and the timing of the RPV failure. In the sensitivity studies the two groups of ruptures were distinguished related to the two issues, what has given the motivation for the further work on these topics. The paper will present the sensitivity study of the corium propagation and will identify the relevance of those two issues for the RPV time rupture.

503

CSN Experience in the Development and Application of a Computer Platform to Verify Consistency of Deterministic and

Probabilistic Arguments in Licensing Safety CasesJose M. Izquierdo1, Javier Hortal2, Enrique Meléndez1, Miguel Sánchez1

1 ConsejodeSeguridadNuclear,CalledePedroJustoDoradoDellmans,11,28040–Madrid,Spain2 NuclearSafetyCouncil,C/JustoDorado11,28040MADRID,[email protected]

This contribution reviews CSN/MOSI activities in the development and application of its own Integrated Safety Assessment (ISA) diagnosis method, designed as a regulatory tool to verify adequate protections of Nuclear Power Plants via a systematic set of analytical tests. The approach harmonizes the probabilistic and deterministic aspects of safety assessments and has been implemented through a consistent, unified and scientifically justified methodology implemented via a computational simulation framework called SCAIS (Simulation Code System for ISA). The document will elaborate on 1) The steps done in the development of the SCAIS software platform and ISA methodology. A brief

description of SCAIS will show some of its most salient capabilities such as:a) Simulation of trees of nuclear accident sequences, where the time evolution following some initiating

event will depend on potential equipment degradations (failures) and occurrence of stochastic phenomena which are systematically explored by the system;

b) Sequence frequency calculations and risk integration taking into account possible dependencies on the plant dynamic state at simulation time step level; and

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c) Simulation of operator actions.2) Recent applications of the tool that illustrate its potentiality in the following assessment aspects: • Completeness of the event Tree delineation;• Emergency Operating Procedures;• Safety Margins; and• PSA Success Criteria, including available times for operator actions.With these developments, CSN/MOSI aims• to show a strong evidence to advocate the development of diagnosis tools/methods specific for TSO

and nuclear safety regulatory bodies tasks;• to argue in favour of a specific research topic (hopefully within an international cooperative effort

among national partners) aimed at developing diagnosis tools and methods that effectively combine probabilistic and deterministic arguments; and

• to contribute to a more objective concept of Risk Informed Decision-Making.

504

Simulation of Hydrogen Combustion Experiment in Large-Scale Experimental Facility with ANSYS Fluent CFD Code

Tadej Holler1, Ed Komen2, Ivo Kljenak1

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 NRG-NuclearResearchandConsultancyGroupDept.Fuels,ActinidesandIsotopes,P.O.Box25, 1755ZGPetten,[email protected]

Possible interaction of steam with an uncovered reactor core during a severe accident in a Light Water Reactor (LWR) Nuclear Power Plant (NPP) can result in generation of large amounts of hydrogen. When hydrogen is released into the containment, a highly flammable gas mixture may form when the hydrogen is mixed with air. The risk of hydrogen deflagration has received increased attention after the Three Mile Island accident in the USA back in 1979, and also most recently following the Fukushima accident in Japan in 2011, where hydrogen destructive power was displayed. Even though hydrogen mitigation systems, like for example passive auto-catalytic recombiners (PARs) and igniters, can be installed in order to reduce the risk of hydrogen combustion as far as possible, theoretical modelling is still required for the optimal design of hydrogen mitigations systems and the assessment of the accompanied residual risks of the presence of hydrogen. Complementary to the use of so-called lumped parameter codes, Computational Fluid Dynamics (CFD) modeling can be used for more detailed assessment of the hydrogen risks in determining the possibility of a breach of the NPP’s containment integrity. For that purpose, extensive validation of CFD codes is a prerequisite.This paper presents validation results of a CFD combustion model using a uniform hydrogen-air-steam mixture deflagration experiment performed in large-scale HYKA A2 experimental facility at the Karlsruhe Institute of Technology in Germany. This experimental facility, with its volume of 220 m3, is considered to be one of the largest available facilities for hydrogen combustion experiments and offers a good midpoint validation benchmark towards the real scale containment.The simulation results of the performed experiment were obtained using ANSYS Fluent CFD code with Flame Speed Closure (FSC) combustion model implemented via user defined functions. This combustion model was developed to simulate combustion not only in fully developed turbulent premixed flames but also in weakly turbulent combustion regimes. The preferential diffusion thermal instabilities (PDT) effects as well as thermal radiation were also accounted for in the applied CFD model.A detailed analysis and comparison with the experimental results of pressure history, including determining maximum pressure with average and maximum pressure increase rate, as well as flame development and propagation, were conducted. These parameters are considered to be of crucial importance in determining the consequences of hydrogen combustion.

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Since the combustion in the HYKA A2 experimental facility took place mostly in a low turbulent regime, the CFD simulation results obtained using the FSC combustion model are expected to match experimental results with reasonable accuracy both qualitatively and quantitatively.

505

Uncertainty and sensitivity analysis of the Generic Containment SB-LOCA accident

Mantas Povilaitis1, St. Kelm2, Egidijus Urbonavičius1

1 LithuanianEnergyInstitute(LEI),LaboratoryofNuclearInstallationSafety,Breslaujos3, LT-44403KAUNAS,Lithuania2 InstituteforSafetyResearchandReactorTechnologyForschungszentrumJuelich,52425Juelich, [email protected]

Within the FP7 SARNET2 project a Generic Containment nodalisation, based on a German PWR (1300 MWel), was developed and used in the benchmark exercise in order to compare simulations performed with various codes using the same nodalisation. Recommendation to elaborate such Generic Containment concept was one of the outcomes of the OECD/NEA ISP-47 activity to allow comparison of the results obtained by different lumped-parameter models on plant scale. Even though the model of the Generic Containment and the transient scenario were precisely and uniquely defined, very different results were obtained not only between different codes but also between participants using the same code, showing that the »user-effect« can have significant influence on the results.The aim of this work is to estimate an uncertainty related to the initial and boundary conditions of the simulated transient, e.g., initial temperature and source term, and compare it to the »user effect« related uncertainty obtained in the SARNET2 project. The paper presents uncertainty estimation using GRS method for uncertainty and sensitivity evaluation and obtained results in the context of the “user-effect” uncertainties.

506

Potential of vapour explosions in sodiumMitja Uršič, Matjaž LeskovarJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

One of the important issues in core melt progression during a severe accident in an innovative sodium cooled fast reactor is the likelihood and the consequences of a vapour explosion. A vapour explosion may occur when the hot core melt comes into contact with the liquid sodium. A strong enough vapour explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.Experiments showed that vapour explosions could occur in sodium. These experiments revealed also an important effect of the sodium sub-cooling on the behaviour of the melt-sodium interaction. The vapour explosion probability and efficiency for a higher sub-cooling is lower than for a lower sub-cooling. The physical properties of sodium, which strongly affect the melt-sodium heat transfer, and the melt solidification, which strongly affects the energy efficiency during the explosion, are identified as the reason for the observed behaviour.Experimentally observed relevant conditions for the vapour explosion in sodium will be simulated in the real experimental geometry. The mixing of the melt with sodium will be simulated with the MC3D

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code (IRSN, France). A sensitivity study of most relevant parameters will be done and the results will be presented and discussed. Based on the performed study the relevance of the sodium physical properties and the melt solidification for the potential of vapour explosion in sodium will be assessed.

507

Probabilistic Safety Assessment of Shutdown and Refueling States

Mitja Antončič, Živa Bricman Rejc, Marko ČepinFakultetazaelektrotehniko,Tržaškacesta25,1000Ljubljana,[email protected]

Probabilistic safety assessment is a standard method for identifying and assessing risk associated with nuclear power plants and other complex technologies for the purpose of improving their safety and performance.The main objective of the work is a probabilistic safety assessment model development and its analysis for cold shutdown and refuelling of a nuclear power plant with pressurized water reactor.The method of work includes definition of several plant operating states, for which the operating conditions, system requirements, plant configuration, initiating events, function events and end states and are determined. Overall risk assessment requires evaluation of risks of separately analysed plant operating states and their time durations. The event trees and the fault trees including the failure criteria of affected systems are developed for each plant operating state. The models are analysed.The model analysis and the results are obtained for each plant operating state. The end states frequencies and the importance factors of components differ for different plant operating states. The results show that the risk of plant shutdown and refuelling states is not negligible.

508

Analysis of LOFA in BWR Spent Fuel Storage PoolCemil Kocar1, Cigdem Polat Dagli2

1 HacettepeUniversity,NuclearEngineeringDepartment,06800Beytepe,Ankara,Turkey2 TurkishAtomicEnergyAuthorityAnkaraNuclearResearchAndTrainingCenter,EskisehirYolu9km, Lodumlu,06530Ankara,[email protected]

After the earthquake and tsunami disaster in Japan, INES-7 scaled accident had happened in Fukushima Daiichi Nuclear Power Plant. When the earthquake and tsunami occurred, Fukushima Unit 4 was in its periodic shut-down stage, and all fuel bundles in the core had been moved to the spent fuel pool. Approximately four days after the station black out, there was an explosion observed at Unit 4, then all focused on the spent fuel storage pool. This study focus on the loss of coolant flow accident in spent fuel storage pool which is modeled by using RELAP5/SCDAP code to observe the coolant level reduction and fuel uncovery because of decay heat of the fuel in the pool.

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509

In vessel melt retention (IVMR) for a VVER 1000 reactor – CFD computation

Aleksander GrahEuropeanCommission,DGJRC,InstituteforEnergyandTransport,P.O.Box2,NL-1755ZGPetten,[email protected]

Since the events in the Fukushima power station studies of station blackout accidents gained interest. One of possible events is a reactor core meltdown due to lack of cooling. The stabilization of molten corium is recognised as essential if a safe and stable state is to be reached following a severe accident. Among the possible options, In-Vessel Melt Retention (IVMR) appears as an attractive solution that would minimize the risks of containment failure (less Hydrogen produced, no corium-concrete interaction), if it can be proved to be feasible. This paper presents first CFD studies with a realistic design of a VVER 1000 reactor including a two-phase corium melt, the reactor pressure vessel wall, and boiling water in the flooded reactor pit. It is assumed that the reactor wall is equipped with a shield along the outer surface of the wall to enhance natural circulation. Heat transfer over the wall is discussed with respect to the critical heat flux.

510

Analysis of ZrO2/WO3 vs. ZrO2/UO2 Fuel-Coolant Interaction in KROTOS Conditions

Vasilij Centrih, Matjaž LeskovarJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

A steam explosion is an energetic fuel-coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. Within the nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations. Stratified melt-coolant configurations, i.e. a molten corium layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in the stratified melt-coolant configuration. In order to validate the explosivity of the melt-coolant interaction in the stratified configuration, the explosion potential of the material that was used in the PULiMS and SES experiments, i.e. eutectic mixture of ZrO2/WO3, should be analysed and estimated also in the conventional melt jet-coolant pool configuration. This can be done by the evaluation of the explosivity of the ZrO2/WO3 melt in conditions of KROTOS experiments, which were performed in the melt jet-coolant pool configuration with prototypic ZrO2/UO2 corium melt.In the paper, the performed analysis of the explosion potential of the PULiMS/SES material ZrO2/WO3 will be presented and discussed. Various simulations were performed with the MC3D code in the KROTOS geometry for OECD SERENA project like conditions. The PULiMS/SES material was compared with the reference prototypic material ZrO2/UO2. The influence of important parameters and typical experimental conditions on the simulation results was analysed, such as the initial melt temperature and the explosion triggering time. The calculations were compared with the simulation of the SERENA KS-4 experiment with ZrO2/UO2 material. Based on the results, the implications of the material influence on the stratified steam explosion energetics and the probability of the spontaneous explosion may be addressed. The calculations in the presented study can be used as a basis for the optional future KROTOS experimental test with the PULiMS/SES material which would support the analytical work.

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511

Fission Product and Aerosol Deposition Analysis in Phebus Containment Under FPT3

Aurimas KontautasLithuanianEnergyInstitute,Breslaujosstr.3,LT-44403Kaunas,[email protected]

The Phebus FP (Fission Product) programme was implemented in between 1988 and 2010 by the “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN, France). The FPT3 test is the last in Phebus FP programme accomplished in 2004. This program was initiated after severe accidents in nuclear reactors at Three Mile Island in 1979 and at Chernobyl in 1986. The Phebus FP program involved a range of experimental and analytical researches from melting of a fuel assembly to release of fission products and structural materials through the primary circuit to the containment atmosphere and a huge database of valuable information was created. Although, a series of experimental and analitical researches has been performed in 30 years, but and nowadays a risk of a hypothetical loss of core cooling in nuclear reactor exist, and the Fukushima Daiichi nuclear disaster in 2011 could be a witness. In this paper is presented continuation of an analitical researches of fission products and aerosols deposition processes in Phebus containment under FPT-3 test conditions. COCOSYS and ASTEC codes were used for the analysis during the core degradation and aerosol phases. Containment model of 16 nodes was used for the analysis, which was already succesfully used for the FPT-1 and FPT-2 tests analysis. Good agreement of thermal-hydraulic parameters results with the test results provides a good basis for further particles deposition analysis. The performed aerosol and fission product analysis mostly confirms previoulsy received results involving created model potentiality to simulate particles deposition processes in containment atmosphere, but also gives and new guidelines concerning fission product deposition distribution.

512

Analyses of THINA melt-sodium interaction experiments with MC3D

Mitja Uršič, Matjaž LeskovarJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

The construction of several demonstration-scale reactors cooled with sodium is planned in the next years. In the frame of safety studies for the demonstration-scale reactors it is important to estimate the risk for the environment in case of a severe accident, including vapour explosions.The modelling capabilities of the fuel-coolant interaction codes to study the vapour explosion phenomenon in light-water reactors were already proven in the frame of the OECD SERENA (Steam Explosion REsolution for Nuclear Applications) and EU SARNET (Severe Accident Research NETwork of Excellence) programmes. Because of the large differences in thermo-dynamical and physical properties of sodium compared to water, the applicability of the fuel-coolant interaction codes and models for the fuel-sodium interaction must be demonstrated. Recently, it was demonstrated that the MC3D code (IRSN, France) has the capabilities to simulate the fuel-sodium interaction.In the frame of the code verification and validation, the applicability of the MC3D code must be further demonstrated. For that purpose, the THINA (KfK, Germany) experiments were simulated. The non-explosive THINA experiments are relevant for the assessment of the fuel-sodium interaction modelling prior the explosion phase. In the paper the experimental and simulation results will be presented, analysed and discussed.

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Radioactive Waste, Environmental Issues


Radioactive Waste, Environmental issues601

Preparation of the national program for the spent fuel and radioactive waste management taking into account

possibility of European Repository Development Organisation development

Tomaž Žagar, Leon KegelARAO–Agencijazaradioaktivneodpadke,Celovškacesta182,1000Ljubljana,[email protected]

According to Council Directive 2011/70/Euratom establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste each European Union member state shall ensure the implementation of its national program for the management of spent fuel (SF) and radioactive waste (RW). On 1 February 2006, the Parliament of the Republic of Slovenia passed the Resolution on the 2006–2015 National Program for Managing Radioactive Waste and Spent Nuclear Fuel (Official Gazette of the Republic of Slovenia, No. 15/2006). Slovenian government is currently reviewing valid national program, which expires at the end of 2015. The document will sets out general timelines and financing for activities related to SF and RW management for all radiation and nuclear facilities in Slovenia for the next ten years (2016-2025) in the fields of low level and high level waste management. Slovenia, like many other countries, operates small nuclear fleet and can be expected to generate relatively small amounts of spent fuel. For such a small program financial and human resources required to develop a national disposal facility may not be feasible or economically practical within the framework of an open and connected markets of European Union. A multinational repository that would accept spent fuel or waste packages from such countries could present a potential solution for disposal challenges. In the European Commission this idea was recognised some years ago and was developed through two SAPIERR preparatory projects and is now being further promoted trough a Working Group on European Repository Development Organization (ERDO-WG). The paper presents an approach how to incorporate this multinational/regional developments in the field of SF management in national program through dual track approach.

602

On-line relative air dispersion concentrations one week forecast for Krško NPP prepared for routine and emergency

usePrimož Mlakar1, Boštjan Grašič1, Marija Zlata Božnar1, Borut Breznik2

1 MEISstoritvezaokoljed.o.o.,MaliVrhpriŠmarju78,1293Šmarje-Sap,Slovenia2 NuklearnaelektrarnaKrško,Vrbina12,8270Krško,[email protected]

Realistic air pollution forecast over complex terrain of Slovenia is a research challenge. A forecast is needed to optimize the timing of routine releases. And on the other hand a forecast is essential in the case of a nuclear emergency to allow for proper action planning. In both cases it is desirable to have as low as

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possible the concentrations at ground level and cloud shine, both resulting as doses to inhabitants and NPP crew.In diagnostic mode relative concentrations are available in on-line mode to Krško NPP crew since 2002. Numerical Lagrangian particle air pollution dispersion modelling system routinely calculates relative concentrations (X/Q) on a half-hourly basis using meteorological measurements (ground level stations and vertical profiles).In the beginning of 2015 the system has been significantly upgraded with forecast of relative concentrations for up to 7 days ahead.Firstly a dedicated fine resolution weather forecast has been developed for the area of Krško NPP and its surroundings. The forecast runs at MEIS cluster. It is based on WRF Weather Research & Forecast model, its ARW (advanced Research WRF) research version that is under development/improvement process in NOAA/NCEP, USA. Double nesting is performed. The time resolution of inner domain of this 7 days weather forecast is half hour and space resolution is 2 kilometres.Based on this weather forecast a forecast of relative concentrations is calculated for the whole forecast period. Relative concentrations are calculated for 5 most probable sources of emissions. These sources are plant ventilation, passive filters exhaust, ejector, ground level release, steam exhaust. Relative concentrations for the forecast for the following 7 days are calculated for the same domain, with the same resolution and with the same numerical Lagrangian particle dispersion model as in diagnostic mode.Relative concentrations, diagnostic or prognostic ones, can then be used by DOZE dose projection system to estimate doses to inhabitants. The whole system of forecasts including basic validation of weather forecast will be described in details.

603

Transfer of Th-230 from soil contaminated with U-mill tailing to radish, savoy and rocket

Petra Planinšek, Borut Smodiš, Ljudmila BenedikJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Uranium mining and milling activities may result in elevated activities of natural radionuclides in the vicinity of facilities. During the operation of the former uranium mine Žirovski vrh, Slovenia in 80s, mining and milling wastes were deposited on nearby waste piles. The wastes contain elevated levels of natural radionuclides from the uranium decay chain, including Th-230. However, radionuclides can be with different migration process (erosion, aerial deposition, groundwater) transferred into the environment and taken up by plants. In this study several edible plants form Brassicaceae family known as plants with higher ability to accumulate heavy metals as well as radionuclides were cultivated. As study plants radish (Raphanus sativus L.), savoy (Brassica oleracea var. sabauda) and rocket (Diplotaxis tenuifolia) were used to assess accumulation of Th-230 radioisotope. The plants were grown in soil mixed up with different shares of waste uranium mill tailings. After growing season, Th-230 in contaminated soil, above ground and underground parts were analysed. Measurements were performed by alpha particle and gamma-ray spectrometry.The obtained activity concentrations of Th-230 in leafs of radish were 1.5 ± 0.5, 1.9 ± 0.4, 1.7 ± 0.7, 2.1 ± 0.8 and 2.0 ± 0.7 Bq/kg of dry mass, and in tubers of radish 2.4 ± 0.7, 2.4 ± 0.8, 3.2 ± 1.1, 3.9 ± 1.0 and 3.4 ± 0.1 Bq/kg of dry mass, for control, 20, 40, 60 and 80 % of uranium mill tailing content, respectively. On the other hand the activity concentrations in leafs of savoy were 1.7 ± 0.5, 1.6 ± 0.4, 1.5 ± 0.3, 2.1 ± 0.7 and 2.3 ± 0.6 Bq/kg of dry mass for control, 20, 40, 60 and 80 % of uranium mill tailing content, respectively. Moreover the activity concentrations in leafs of rocket were 2.4 ± 0.7, 4.0 ± 2.0, 3.0 ± 0.7, 4.5 ± 1.8 and 7.5 ± 0.2 Bq/kg of dry mass for control, 20, 40, 60 and 80 % of uranium mill tailing content, respectively.The results obtained showed that no correlation among Th-230 activity concentrations and different portions of mill tailing were observed for radish and savoy. However, in case of rocket the lowest

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Th-230 activity concentration was measured for the control sample and the highest for plants growing in substrate with 80 % of uranium mill content. The experiment showed that the most promising candidate for phytoremediation of areas contaminated with thorium isotopes can be rocket. Nevertheless, further investigations are needed to study influence of other growing conditions like soil type and climate in more detail.

605

Barriers and Operational Risk Assessment of Incidents and Accidents occurring in the Transport of Radioactive Materials

Thomas Breznik, Marko Gerbec, Borut SmodišJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Investigations of major incidents and accidents in the transport of radioactive materials (TRAM) show that technical, human, operational, as well as organisational factors influence the incident rate sequences in everyday TRAM performance. In spite of these facts, quantitative or semi-quantitative risk analyses of human error beside technical factor as main cause of all emergency events in the transport chain should be performed. Therefore, the main focus should be on generation, performance and evaluation of human organisational systems such as generic safety barriers, failure modes and their control measures. The paper presents a method called Barrier and Operational Risk Analysis (BORA-TRAM) in the qualitative and semi-quantitative risk analysis in the general performance of TRAM. Application of BORA-TRAM for analysis of the loss of containment barrier evidently presents a more detailed risk picture than the traditional quantitative risk analyses (QRA), since no analyses of causal factors of RAM release are carried out in the existing QRA.By using BORA-TRAM it is possible to analyse the effect of safety barriers introduced to prevent loss of containment and control (radiation releases and radiation exposition). Furthermore, it reveals how platform specific conditions of technical, human, operational, and organisational risk influencing factors affect the barrier performance in final risk evaluation of TRAM. BORA-TRAM comprises the following main steps: 1) System Identification (development of a basic risk model including release scenarios; this is often done

by task analysis);2) Modelling the performance of safety barriers (literature review). This is the key point that can incorporate

human organizational and operational risk influencing factors (RIFs) into the barriers and then into the initiating events;

3) Assignment of generic data and risk quantification based on these data; establishment of a fault tree. This is used for analysis of barrier performance. All of basic events in TRAM should be analyzed;

4) Development of risk influence diagrams. This should cover representative scenarios, and it usually consists of initiating event, barriers and outcomes. This barrier block diagram can also be converted into an event tree;

5) Scoring of RIFs;6) Weighting of RIFs;7) Adjustment of generic input data;8) Recalculation of the risk in order to determine the platform specific risk related to radioactive material-

RAM package loss or release. The various steps in BORA-TRAM will be presented and discussed. Final results from a case study where BORA-TRAM is applied and revision of the proposed method in risk assessment in TRAM will also be presented.

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606

Radiochemical techniques for determination of actinides, Po-210, H-3, C-14 and Sr-89/90 in urine samples

Ljudmila Benedik, Marko Štrok, Barbara Svetek, Zdenka TrkovJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Department of Environmental Sciences at the Jožef Stefan Institute is the only laboratory in Slovenia that provides analysis of alpha emitters in various samples. It is accredited for determination of H-3, C-14, Sr-89/90 in environmental samples. In order to assure the quality of analytical results in recent years participated in numerous intercomparision exercises for determination of natural and man-made radionuclides organized by international organization such as the International Atomic Energy Agency, the Institute for Reference Materials and Measurements, the National Physical Laboratory, the Bundesamt für Strahlenschutz, Procorad and others.Monitoring of internal exposure to radiation for workers at a nuclear facilities is an important part of radiation protection system. PROCORAD (Association for the Promotion of Quality Control in Radiotoxicological Analysis) organises radiotoxicology intercomparison in order to evaluate the quality of medical analysis results and to promote good laboratory practice. Yearly the Department of Environmental Sciences participates in this intercomparison. Urine samples, provided by the organizers, are real biological samples which contain the radionuclides of occupational exposure workers.In the present work radiochemical procedures for determination of actinides (thorium and uranium radioisotopes, Pu-238 and Pu-239+240, Am-241), Po-210, H-3, C-14 and Sr-89/90 in urine samples were shown. Actinides (Th-228, Th-230, Th-232, U-234, U-235, U-238, Pu-238, Pu-239+240, Am-241) were determined by alpha-particle spectrometry. After addition of radioactive tracers (Th-229, U-232, Pu-242 and Am-243), samples were evaporated to dryness. After wet ashing with mineral acids, radiochemical separations by ion-exchange chromatography and extraction chromatography for investigated radionuclides were performed. Source preparation for alpha-particle spectrometry was carried out by the micro-coprecipitation method with neodymium fluoride.Similar to actinides, Po-210 was also determined by alpha-particle spectrometry. After addition of radioactive tracer Po-209, the sample was evaporated to dryness and digested with mineral acids. Afterwards, polonium was deposited on Ag disc and measured by alpha-particle spectrometry.Samples for tritium analyses were distilled and tritium was determined using liquid scintillation counting with the method of standard addition. Ultima Gold LLT was used as scintillation cocktail.In the case of C-14, samples were mixed with scintillation cocktail Ultima Gold LLT and measured with liquid scintillation counter with the method of standard addition.For Sr-89/90, stable Sr carrier was added in order to gravimetrically determine radiochemical recovery. Sr-89/90 was separated from other radionuclides with subsequent precipitation of strontium oxalate, nitrate, iron hydroxide and barium chromate. Finally, Sr was precipitated in the form of strontium carbonate, transferred to counting planchet and measured with low background gas flow proportional counter.

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608

Radioanalytical methods laboratory – past, present and futureL. T. Dobrev, B. Slavchev, A. ChalakovInstituteforNuclearResearchandNuclearEnergy,72Tzarigradskochaussee,Blvd.,BG-1784Sofia,[email protected]

Radioanalytical methods laboratory was created in the middle of last century, in the year of 1969 as a part of IRT-2000 of Physical Institute of BAS. Until 1989, laboratory’s field of interest was scientific researches in area of instrumental and radiochemical neutron-activation analysis. As a part of INRNE-BAS, the laboratory is specialized in the field of radioanalytical methods for determination of radioactive wastes and environmental samples. In the last years, Radioanalytical methods laboratory is focused extremely on determination of radionuclide content of special type liquid radioactive waste, called evaporation concentrate. The laboratory was developed various number of complex radioanalytical procedures for analysis of solid crystal and liquid phase of the evaporation concentrate. The development of laboratory is appointed in expanding of activities and range of used instrumental methods such as alpha- and gamma-spectrometry, liquid scintillation counting, gas-chromatography, ICP-MS etc. The laboratory became a leader in the area of complex analysis. Together with radionuclide determination, we can provide and standard chemical analysis of the samples. The mixture of chemical, physical and radioanalytical methods for analysis inserts in the laboratory and INRNE at all, extremely actual and discussed in world scale theme for interdisciplinarity in science. Because of that, Radioanalytical methods became a reference laboratory of nuclear regulatory agency (NRA) – Bulgaria.ACTIVITIES:• Development of radioanalytical methods;• Radioactive samples analysis;• Preparation of secondary standard sources, used for efficiency calibration of alpha and gamma

spectrometry – organic matrix with U-natural.;• Participation in project for evaluation of the fluence through critical welds of the reactor vessels in NPP-

Kozloduy;• Development of methods for determination of α-, β- and γ-emitters in samples from nuclear facilities;• Radioactive waste analysis;• Development of techniques for decontamination of metal radioactive waste;• Measurements of radioactive releases from Kozloduy NPP;• Environmental monitoring;

Keywords: Radioanalytical chemistry, laboratory, analysis, INRNE

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609

Breakup and Solidification Behaviour of Liquid Metal Jet in Water Environment

Radu Secareanu1, Minoru Takahashi2, Riccardo Mereu3, Ilie Prisecaru1

1 UniversityPolitehnicaofBucharest,312,SplaiulIndependentei,Bucharest,Romania2 TokyoInstituteofTechnology,ResearchLaboratoryforNuclearReactors,2-1Hirosawa,Wako-shi, Saitama,351-0198,Japan3 PolitecnicodiMilano,DipartimentodiEnergia,ViaPonzio34/3,20133Milano,[email protected]

The study was motivated by the recent activities for the decommissioning of Fukushima Daiichi Nuclear Power Plants. In order to remove the melted core from the reactor pressure vessels or reactor containment vessel, it is necessary to fill the last one with water to act as a radiation shield. At a preliminary test numerous leak points were discovered in the drywell, suppression pool and torus room. Based on the Fukushima decommissioning roadmap it is necessary to develop a new technology to stop the water leak with a seal that must resist for 30-40 years during reactor decommissioning process. The present study comes with an alternative sealing method by using liquid metal. The purpose is to use a low temperature melting metal to solidify near the leak surface inside the suppression pool. The main advantage of this ideea is the posibility to control the freezing and to reduce the solidification time. The promising candidate of the liquid metal may be lead alloys, particularly Wood’s metal, since it’s melting point is less than 100 oC. The low melting point is important in order to prevent local boiling of the water which can influence the injection. Lead alloy technology such as lead-bismuth eutectic has been developed for a long time since the lead alloys have been proposed as coolant for fast reactors and they are also efficient as a radiation shield. Because of its stable properties under radiation, good sealing condition by the use of lead alloy can be obtained for long periods of time.The present study is experimental and is measureing the jet temperature in order to observe the freezing behaviour. Also with the help of a fast camera the jet breakup and the responsable instabilities are captured and analized for each breakup mechanism with the help of ambient Weber number. The understanding of the jet breakup and freezing characteristics is important in order to develop a sealing technology based on liquid metal injection in water, and in the end of the paper the link between the instabilities, breakup mechanism and jet temperature history is showed.

610

Traveller Implementation and Experiences at NPP KrškoMarko Gordić, Dejvi Kadivnik, Bojan Kurinčič, Martin ChambersNuklearnaelektrarnaKrško,Vrbina12,8270Krško,[email protected]

Traveller is a shipping package designed to transport non-irradiated uranium fuel assemblies or rods with enrichments up to 5.0 weight percent. It is designed for transport of several types of PWR fuel assemblies as well as either BWR or PWR rods and can carry one (1) fuel assembly or one (1) pipe for loose rods. The Traveller was designed and manufactured to comply fully with the requirements of US regulations for safe Packaging and Transportation of Radioactive (10CFR711) as well as IAEA Regulations for the Safe Transport of Radioactive Material (TS-R-1). NEK introduced the Traveller shipping package as a replacement for aging MCC-3 shipping containers, which were designed in the 1970s and as such do not comply with contemporary international transport regulations. They are also not manufactured anymore with existing containers having to be periodically

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refurbished and maintained.Upon receiving necessary licencing approval in 2014, NEK transitioned to the Travellers and successfully performed the first fresh fuel receipt in late 2014 using new shipping package design. Experiences during the implementation and fuel receipt are described and compared to the previous design with MCC-3 shipping containers.

611

Radiation Protection Training Needs in SloveniaMatjaž Koželj, Igor JenčičJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Slovenia has three nuclear facilities, more than 200 industrial and research organisations with nearly 1,000 X-ray and radioactive sources, almost 800 dental, diagnostic and other medical X-ray devices in numerous medical institution. In some medical and research institutions unsealed sources are also being used. There are also radiation practices where considerable number of workers are exposed to natural sources, like radon or cosmic radiation. The number of radiation workers, i.e. people working with artificial sources, or being exposed to elevated levels of radiation from natural sources in Slovenia exceeds 6,000 in the last few years, and it is still increasing.Radiation workers, and also workers professionally involved in radiation protection implementation in nuclear facilities and licenced organisations as radiation protection officers must be adequately trained in radiation protection. The content and duration of radiation protection training for different groups of radiation workers is defined in “Rules on the obligations of the person carrying out a radiation practice and person possessing an ionizing radiation source” where 17 different programs for different profiles of radiation workers (and also for radiation protection officers) are defined.Annually, more than 1,500 workers attend radiation protection courses organised by three organisations with approved programs. While most of workers belong to a few profiles, there is also irregular but frequent need for courses for other profiles. Since the implementation for these courses must also be in accordance with authorised programs, it requires not only careful planning but also involvement of additional resources, which increases total costs of training.This problem, i.e. the complexity of profiles is evident especially for radiation protection officers. The required training for radiation protection officers ranges from 10 to 200 hours. The longest one is intended for members of radiation protection unit staff in nuclear facilities and assumes work and efforts of many people. Considering all these facts, it is our intention to define realistic needs for radiation protection training regarding profiles and number of workers and to suggest some optimisation of required training scheme. This also involves implementation of new requirements from new European BSS which must be implemented in Slovenian legislation until the 2018.

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Nuclear Fusion


Nuclear Fusion701

Deuterium retention studies in self-ion damaged tungsten exposed to neutral atoms

Sabina Markelj1, Anže Založnik1, Thomas Schwarz-Selinger2, Mitja Kelemen1, Primož Vavpetič1, Primož Pelicon1

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 Max-Planck-InstitutfürPlasmaphysik(IPP),Boltzmannstr.2,D-85748Garching,[email protected]

For the prediction of tritium retention in future fusion reactors (ITER and DEMO) influence of irradiation by neutrons produced by the fusion reaction D-T has to be taken into account. In that respect, successful development of neutron resistant materials needs to be made. Present indications show that tungsten is the most suitable ‘baseline’ material as plasma facing component armour (from the EFDA-roadmap). Since neutron irradiated samples are activated and therefore need special handling, high energy ions are used as neutron surrogates to produce neutron-like damaged material. It was shown that fuel retention in damaged tungsten is strongly increased as compared to undamaged tungsten material and for this reason influence of exposure temperature and damage annealing is being extensively studied [1, 2, 3]. In order to study deuterium retention in damaged tungsten, the samples were damaged by 20 MeV tungsten ions at room temperature resulting in a 2.4 µm deep damaged layer, producing defects similar to those created by neutrons. In order to study the extent of neutron-like damage being produced in tungsten material the damaged samples are being exposed to deuterium atoms. The atoms are trapped at the produced defects that act as additional strong binding sites for deuterium atoms in the bulk, without producing any additional damage. We have performed two kinds of experiments with different exposure and sample tempering procedure. In the first case the so called self-ion damaged tungsten samples were exposed to D atoms, flux of 3.5×1019 D/m2s, at sample temperatures from 500 K to 900 K to fluences that were enough to saturate the damaged layer by deuterium atoms. In the other case the samples were first annealed for one hour at different sample temperatures (600 K – 1200 K) and then exposed to D atoms at 500 K to a fluence of 1.3×1025 D/m2. By analysing the deuterium depth profile by Nuclear Reaction Analysis in such treated samples we have studied the effect of sample exposure temperature and sample tempering on damage annealing and deuterium retention. It was found that both exposure and annealing temperature have strong influence on the maximum deuterium concentration and on the integrated amount of deuterium in the sample. In both cases the concentration is decreasing with increase of the temperature. In the case of only defect annealing the deuterium retention is reduced by 70 % at 1200 K as compared to the un-annealed sample. Exposing samples at 900 K retention decreased by 90 % when comparing to exposures at 500 K. For the deuterium exposure at elevated temperatures two processes are included, defect annealing and deuterium de-trapping from traps produced by damaging. From the observed results it was shown that the last one is dominating retention at elevated temperatures.

[1] O. V. Ogorodnikova et al., J. Nucl. Mater. 451, 379 (2014)[2] E. Markina et al., J. Nucl. Mater. (2014)[3] M.H. ‘t Hoen et al Phys. Rev. Lett. 111, 225001-1 (2013)This work has been carried out within the framework of the EUROfusion Consortium, under work program PFC, and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.”

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Nuclear Fusion


702

Ball-pen probe diagnostics of a weakly magnetized discharge plasma column

Lino Šalamon1, Gabrijela Ikovic2, Jernej Kovačič1

1 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 Fakultetazamatematikoinfiziko,Jadranska19,1111Ljubljana,[email protected]

One of the simplest and most widely used probes in low temperature plasmas is the Langmuir probe. It is constructed of a simple and small collector, most commonly from a cylindrical piece of wire. To obtain plasma potential values, the following simplistic relation can be used in Maxwellian plasmas: fi_p=fi_f-(kT/e)lnR, where R=Isat+/Isat-. These procedure requires post processing of data obtained from I-V curve. Ball-pen probe (BPP) has recently been developed for direct measurements of fi_p. Idea is to reduce the Isat- until it reaches Isat+ using a retractable collector, shielded inside an insulating tube. In magnetized plasmas Isat- is screened off due to smaller Larmor radius of electrons comparing to ions until at some depth it decreases down to the Isat+. Ball-pen probe has already been extensively tested and used in tokamak plasmas, however not many studies had been done on the behavior of a BPP in low-temperature plasmas, where EEDF can be far from Maxwellian.Experiments were performed in a linear magnetized plasma device composed of a stainless steel tube 1.5 m long and 17 cm inner diameter with a hot cathode electron source. Plasma is confined by a solenoid magnetic field coils, which can create axial magnetic field up to 0.4 T. Measurements were performed over a broad range of gas pressure and magnetic field densities. Probe is radially movable and the position of the collector inside the insulator tube is set by a computer controlled stepper motor. We investigated the dependence of fi_f of the BPP on the depth of collector insertion for different pressures and magnetic field densities. It appears that the floating potential of the BPP is quite sensitive to the presence of highly energetic primary electrons. For low gas pressures a local minimum in the BPP floating potential versus collector position has been identified. Also, at high pressure ball-pen probe was used inside anodic plasma and in the vicinity of a strong double layer where ion or electron beams are present.

703

Global Thermal Analysis of Demo TokamakOriol Costa Garrido1, Boštjan Končar1, Samo Košmrlj1, Christian Bachmann2, Botond Meszaros2

1JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2 EuropeanFusionDevelopmentAgreementMax-Planck-InstitutfuerPlasmaphysik,Boltzmannstr.2, D-85748GarchingbeiMuenchen,[email protected]

A roadmap to the realisation of fusion energy [1] has been developed in November 2012 with a strategic vision to demonstrate the generation of electrical power by a Demonstration Fusion Power Plant (DEMO) by 2050. The goal-oriented approach of the roadmap is carried out within the common European framework program “EUROfusion” that has been launched in 2014. This is a huge but a very coherent program that addresses all key issues in order to achieve the ultimate goal to produce electricity by the means of fusion energy.Development of an integrated DEMO design is one of the important missions in this context. Analysis of DEMO requirements, system modelling and design integration of the various systems that form the overall DEMO plant have to be addressed taking into account different aspects. The thermal behaviour

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Nuclear Fusion


of the overall tokamak system is being investigated at the Jožef Stefan Institute. The DEMO tokamak is composed of systems operating at very different temperatures, which in addition vary depending on the operational state. The relative thermal expansions/contractions as well as the heating/cooling power exchange between different tokamak systems (in-vessel components, magnets and thermal shields) need to be determined to ensure a consistent tokamak design.In this study a global thermal analysis of DEMO tokamak is performed. Based on the updated CAD design of in-vessel components and magnet system, thermal expansion and thermal radiation analyses of the overall system have been carried out. Magnet thermal shields and the cryostat vessel have been designed additionally to complement the overall DEMO CAD design model. The structural code ABAQUS was used to perform the two separate analyses. The thermal expansion analysis predicts relative displacements of the model components at operational temperatures with respect to the initial room temperature. The results provide essential information on thermal expansion of different components that helped to improve the initial design of the vacuum vessel and toroidal field coils. The main objective of the thermal radiation analysis is however to estimate the heat exchange between the tokamak systems and to predict the power needed to cool down the superconducting magnets. The heat radiated to the superconducting magnets is strongly reduced to about 1 kW by the presence of the thermal shields, which absorb approximately 0.5 MW.

[1] Fusion Electricity – A roadmap to the realisation of fusion energy, F. Romanelli et al., EFDA, 2012.

704

Visualisation of Fusion Related Models Stored in General Grid Description

Leon Kos, Janez Krek, Marijo TelentaUniversityofLjubljana,FacultyofMechanicalEngineering,LECADLaboratory,Aškerčevacesta.6,1000Ljubljana,[email protected]

Efforts for integrated modelling with computer simulations of fusion experiments is increasing each year and request for general description of model grids for various devices was needed. The General Grid Description (GGD), in use within the Code Development for integrated modelling EUROfusion project, represents a general solution for saving grids from various experimental and simulation devices in an easy-to-access solution. Saving and retrieving stored grid data is easy for both experiments and simulations as many currently used codes include support for GGD and underlying storage systems. Visualisation of models in general simulation “workflow” presents one of last steps before simulation is carried out and also presents a basis for post-processing. Work presented in this paper was done in visualisation of results stored in database of Consistent Physical Objects (CPO), widely used in fusion related experiments and simulations by using ParaView as a framework for displaying models. All visualisation of stored grids is done inside newly developed plug-ins for ParaView, thus using the full advantage of ParaView as a framework for user interface and interaction. Grids are “loaded” into ParaView using ordinary “selectors” for addressing specific data in CPO database: machine, user, run and shot numbers to precisely define individual experimental or simulation data. With utilisation of ParaView’s support for saving states (saved loaded plug-ins and layout of their windows, output options, etc.), one can use new visualisation plug-in to produce post-processing results (images) in Unix shell and speed the post-processing in large cases or cases involving parameter scanning.

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Nuclear Fusion


705

Sub-Divertor Neutral Gas Analysis at JET with the ITER-like Wall

Aleksander Drenik1, Martin Oberkofler2, Daniel Alegre3, Uron Kruezi4, Sebastijan Brezinsek5, Marco Wischmeier2, Carine Giroud4, Miran Mozetič1 and JET Contributors*6

1JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia2Max-Planck-InstitutfürPlasmaphysik,85748Garchingb.München,Germany3As.CIEMAT,Av.Complutense40,28040Madrid,Spain4CulhamCentreforFusionEnergy,Abingdon,Oxon,OX143DB,UnitedKingdom5ForschungszentrumJülichGmbH,InstitutfürEnergie-undKlimaforschung-Plasmaphysik,52425Jülich, Germany6 EUROfusionConsortium,JET,CulhamScienceCentre,OX143DB,Abingdon,[email protected]

Since the installation of the ITER-like wall, JET offers a unique opportunity to address plasma-wall interaction issues in the same plasma-facing material configuration as will be implemented in ITER: beryllium as the first wall material and tungsten plasma-facing components in the divertor. Neutral gas analysis provides an important insight into plasma wall interactions by identifying impurities in the exhaust of the fusion device. Beside the impurities that are either formed in the reactor vessel or introduced through leaks, considerable amounts of nitrogen, neon and argon are injected on purpose as radiating impurities. Moreover, argon is routinely used as a component in the disruption mitigation gas. The introduced impurities can be retained on the walls, and released in subsequent discharges, or can react with hydrogen species and form stable molecules, as in the case of formation of ammonia during nitrogen seeded discharges.The presented contribution is based on the analysis of data acquired by a sub-divertor mass spectrometry system at JET. The system consists of a Hiden Analytical HAL 201 RC mass spectrometer (MS), located below the divertor cryo pump. The MS is protected by magnetic shielding, which allows it to provide reliable data throughout all phases of the reactor operation. In the discharge phase, the MS is set to a fast acquisition mode, sampling signal intensities at a selected number of discrete masses, with a reduced sensitivity. The fast acquisition rate of 1.4 seconds per scan allows for recording several data points during the flat-top phase. However, as the instrument has not been calibrated, the results indicate only relative trends, rather than absolute values.The majority of data comes from two relatively un-interrupted sequences of pulses: 85174 – 85456, and 87062 – 87450. This set consists of approximately 400 non-seeded discharges, as well as a smaller number of N2 and Ne seeded discharges.In a deuterium-dominated gas mixture, the isotopologues of all three of the main hydrogen containing impurities (water, methane and ammonia) populate the 16 – 20 AMU range in the recorded spectra. A statistical model is used to obtain their respective abundances from the recorded intensities.In non-seeded discharges, the most prominent impurity is methane, followed by water and nitrogen. In seeded discharges, the seeded impurity usually dominates the content. In nitrogen seeded discharges, significant amounts of ammonia are detected as well. Moreover, seeded nitrogen exhibits a slight legacy effect that lasts up to around 6 discharges. The content of methane, and nitrogen in non-seeded discharges, shows dependence of time-integrated auxiliary heating power, and in some cases the strike point position, indicating that these species are produced through plasma-wall interaction. The content of water, on the other hand, seems to be linked more to the quantity of injected gas. The quantity of ammonia is strongly linked to the amount of injected nitrogen, reaffirming the proposed route of in-vessel ammonia production.

* See the Appendix of F. Romanelli et al., Proceedings of the 25th IAEA Fusion Energy Conference 2014, Saint Petersburg, Russia

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706

Study of some atomic and molecular processes relevant to the tokamak edge plasma modelling

Iztok Čadež, Sabina Markelj, Anže ZaložnikJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

Study of atomic and molecular (A&M) processes involving appropriate particles is a collateral activity for modelling of various technological plasmas. Some of such processes might have key role in the observed phenomena and some are of minor importance. New data for specific processes are collected in the data bases [e.g. 1] in formats convenient for direct importation into modelling codes. Study of binary collision processes as well as processes on the surfaces of the constituents of fusion edge plasma and plasma in negative hydrogen ion sources are of particular importance for fusion development. This includes main constituents, hydrogen and helium but also various impurities such as carbon, oxygen, nitrogen, beryllium, tungsten.A&M processes relevant to the fusion development which are of our interest are those involving hydrogen molecules (H2 and its isotopologues) which are either product of reaction (atom recombination on the surface, dissociation of polyatomic molecules containing hydrogen) or incident particle in the reaction (electron scattering, surface reactions). Dissociative electron attachment in hydrogen (e+H2→H-+H) is of particular importance to us due to its additional relevance for specific vibrational spectroscopy of hydrogen molecules used in our laboratory. In this contribution we will present a detailed discussion of the dissociative electron attachment in hydrogen molecule proceeding through the 2Σg

+ resonant state at around 14eV energy of the incident electron. Products of this process are H- and H-atom in n=2 excited electronic state. Simple local complex potential model for this process does not reproduce the experimental data and this discrepancy will be discussed and explained and results of an improved simple model will be presented.

[1] http://www.adas.ac.uk/; http://molat.obspm.fr/; http://www.hydkin.de/.

707

Interfacing of CAD models to a Common Fusion Modelling Grid Description

Marijo Telenta1, Leon Kos1, Robert Akers2

1 UniversityofLjubljana,FacultyofMechanicalEngineering,LECADLaboratory,Aškerčevacesta.6, 1000Ljubljana,Slovenia2 EUROfusionConsortium,JET,CulhamScienceCentre,OX143DB,Abingdon,[email protected]

Workflow for scientific visualization of a CAD model requires a mesh format that can be read by general visualisation tools such as ParaView or VisIt. In addition, the mesh format is favoured to be stored in a common fusion modelling grid description. The proposed workflow consists of de-featuring the CAD model, meshing it into triangular mesh using PythonOCC library, and storing the mesh in EUROfusion Integrated Modelling (EU-IM) database using General Grid Description (GGD) for a sole purpose of visualisation of the CAD model together with results from various fusion modelling codes. GGD is primarily used by EUROfusion task force for integrated modelling (WPCD) as a common grid description for fusion modelling codes that interchange results on top of the EU-IM database. In order to store the CAD model, OpenCASCADE CAD kernel is used to transform the CAD model into triangular mesh suitable for storage in GGD. Ultimate goal is controlled de-featuring of the CAD model applied to desired level of detail (LOD)

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Nuclear Fusion


such that the model could be used as in input and meshed correctly by different fusion modelling. In this paper, review of procedures for CAD de-featuring is discussed. CAD de-featuring consists of enveloping, removing and simplifying unnecessary details. Also, description of the GGD for visualisation and reuse by fusion modelling codes is presented. PythonOCC library is used to scripts the OpenCASCADE kernel in Python programming language. For CAD data exchange, IGES and STEP standards are used and can serve as a common data format for input to meshing codes. If a “mesher” is unable to read the standard CAD format than this usually means that custom input needs to be prepared by a CAD kernel. The goal of the paper is to interface the CAD data to various meshing tools in use within EUROfusion. Meshing is then treated as a black box that outputs mesh/grid as an input for physics codes. Output mesh, as well results of the codes on these grids are stored in a scientific database format. Ideally, all results should be able to be stored in GGD as this enhances further processing and analysis by visualization tools that are available within WPCD. Getting data into GGD format into database EU-ITM means conversion of the results from other compatible formats. The aim of the project presented is to automate the tedious task of CAD model de-featuring and meshing which is often done manually and prepare it as a CAD service for use in scientific workflow engines to assure provenance.

708

Simulations of an Ion Energy Analyzer using PIC techniqueGabrijela Ikovic1, Lino Šalamon2, Tomaž Gyergyek3

1 Fakultetazamatematikoinfiziko,Jadranska19,1111Ljubljana,Slovenia2 JožefStefanInstitute,Jamovacesta39,1000Ljubljana,Slovenia3 UniversityofLjubljana,FacultyofElectricalEngineering,Tržaška25,1000Ljubljana,[email protected]

The code XPDP1 simulates a bounded plasma within planar electrodes and an external circuit in one dimension. The code uses Particle-in-Cell technique for simulating ions and electrons, the leap-frog method for integrating motion and field equations and Monte Carlo collision model for simulating collisions between charged and neutral particles.An Ion Energy Analyzer is a simple probe consisting of parallel grids and a collector plate behind them and is a very standard diagnostic tool for edge plasmas of fusion devices. It is used in plasma devices for obtaining ion plasma parameters. XPDP1 was modified in order to simulate the Ion Energy Analyzer. The new version includes two conductive grids. The first grid is set to be on a floating potential to repel electrons. The second grid is a discriminator. The potential on this grid is directly driven and only the ions with high enough energy can pass through the grid. The potential on the collector should be low enough to repulse the remaining high-energy electrons.The simulations are used for researching an impact of the analyzer on a plasma. The most important parameter is the distance between the grids. If the distance is too large, the maximum potential between the grids exceeds the potential on the discriminator grid due to the excess of the electrons between the grids. This limit distance varies depending on the opacity of the grid, the initial temperature of the ions and other characteristics of the probe. In the second part of the research the secondary emission of electrons was included to the code. It can be seen from the results of the simulations that the secondary electrons disrupt the measurements. In addition, a third grid was included to the code to show the difference between different types of the probes.

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Materials


Materials801

Strain gradient crystal plasticity approach to modelling micro-plastic flow and localisation in polycrystalline materials

Igor Simonovski1, Tuncay Yalcinkaya2

1 EuropeanCommission,DGJRC,InstituteforEnergyandTransport,P.O.Box2,NL-1755ZGPetten, Netherlands2 MiddleEastTechnicalUniversityDepartmentofAerospaceEngineering,Ankara,[email protected]

Structural materials in the reactor pressure vessels are exposed to a harsh environment, resulting in a number of material degradation processes. Irradiation generates a number of point defects in the atomic structure of a material. In addition, plastic slip localization occurs on the grain level size where highly-deformed narrow bands of material appear already at the moderate strain levels. These bands are called channels or clear bands, because they are almost empty of irradiation defects, whereas the surrounding matrix is still full of them. Clear bands are very thin with a thickness of a few tens of nm. It is thought that these clear bands contribute significantly to yield-stress increase and loss of work-hardening and ductility under irradiation. Additionally, high stress concentrations are generated at points where clear bands impinge on the grain boundaries, resulting in grain boundary damage and increasing the possibility of intergranular cracking. Continuum-based structural-mechanics models are not able to predict the initiation or the evolution of grain-level plastic slip localization. New approaches like strain-gradient crystal plasticity are being developed to tackle these issues. In the present work a numerical approach is presented where the application of strain-gradient crystal plasticity is extended to aggregates containing up to tens of grains. Plastic slip localization is demonstrated within the corresponding finite-element model.

802

High Temperature Pipe Structural Health Monitoring System utilising Phased Array probes on TOFD configuration

Ivan Vican1, Channa Nageswaran2, Nikos Makris3, Alvaro Garcia4, Abbas Mohimi5, Stephan Michau6, Marko Budimir1

1 INETEC-InstituteforNuclearTechnology,Dolenica28,10250Zagreb,Croatia2 TWILtd,UnitedKingdom3 IknowHowLtd,Greece4 InnoTecUKLtd,UnitedKingdom5 BrunelInnovationCentre,BrunelUniversity,London,UnitedKingdom6 VermontYankeeNuclearPowerCorp.,FerryRoad,BrattleboroVermont05301,[email protected]

High temperature pipe cracks are the root of a steam power failure in the EU typically every 4 years, resulting in loss of human life, serious accidents, widespread power cuts and massive financial losses for the operators. According to IAEA’s Reference Technology Database such an event on a nuclear power plant has an average cost of €120 million, including outage costs, emergency repair costs, insurance and legal costs. Since only one growing crack is needed to cause a major failure, they have to be inspected and monitored thoroughly.

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Materials


Breakdowns at these extreme conditions (580°C, 400bar) are a result of two major weld failure modes: a) creep cracks near pipe welds: as the high pressures produce a constant hoop stress on the full length of the pipe and high temperatures increase creep deformation; b) fatigue cracks on pipe welds: as vibrations produce cyclic stresses that lead to fatigue type damage. The most prone to defects areas include thick section pipes and headers, which are mostly composed of ferritic steels. Particularly, all over the world the most common material for steam pipes and headers is the P91 steel, primarily employed for T≤593°C. In Europe for steam temperatures up to 620°C, E911 steel is used as it presents increased high-T strength capabilities. Beyond 620°C, 12% Cr steels are employed with HT91 steel being the most popular choice in European power plants.Current maintenance practice is to proceed with repairs on a detected crack according to its severity. For cost reasons, cracks that are not judged as severe enough will not be repaired. Crack severity judgement is based on its probability to cause a failure and this probability is derived taking into account the crack size and operational lifetime. More variables such as operating temperature and vibrations may rarely be found in other studies. Recent data from fracture mechanical statistical studies shows this connection between the size of a crack on a nuclear power plant pipe and its probability to lead to a failure.To deal with the above problems the Structural Health Monitoring (SHM) system is developed and presented in this work, able to achieve continuous operation for a long period, specifically designed for high temperature, high pressure pipelines. The developed system will employ novel Phased Array (PA) ultrasonic probes able to withstand and continuously operate at 580°C. The system is designed to be permanently mounted on superheated steam pipes, at locations of known defects and it continuously monitors their size. However, this supposes that defects will have already been detected by a traditional method during an outage, thus the insulation will have already been removed. The PA transducers are placed according to the Time Of Flight Diffraction (TOFD) technique’s topology, thus creating a novel configuration. Particularly, the high-T transducers are be placed on either side of the weld where the crack being monitored is located. This can enable the system to continuously track crack growth with high accuracy, enabling maintenance crews to estimate the severity directly and not through statistics.

803

Characterisation of coatings evaluated for LFR applicationsFosca Di Gabriele1, Alessandro Gessi2, Petra Bublikova1, Hana Jirkova1, Dagmar Bublikova1

1 ResearchcentreRez,Hlavni130,25068Husinec-Řež,CzechRepublic2 ENEA,ViaMartiridiMonteSole4,40129Bologna,[email protected]

Heavy liquid metals (HLM), such as Lead, Pb, are considered among the possible coolants in the development of Generation IV fast Reactors. One of the critical issues to be solved for the successful development of such a concept, is the interaction of structural materials with LM. Both austenitic and ferritic-martensitic steels are considered and are susceptible to damage (dissolution of alloying elements in the LM) and/or degradation of mechanical properties. One of the candidate materials for Lead Fast Reactors is the austenitic steel 15-15Ti. However, at high temperature also this material may suffer damage, due to Ni dissolution in the coolant. Therefore, one of the approaches to mitigate the damage due to dissolution of alloying elements is the development of stable and protective scales. However, to promote stability and reduce the amount of oxygen in the LM, the use of coatings was proposed. This are layers deposited on the material surface, which have a different chemical composition compared to the base material and works as a barrier against the effect of LM. Coatings are proposed as a valid protection against high temperature damage in this environment. They can work on the principle of facilitating the growth and stability of protective oxides, by introducing oxide forming elements (e.g. Al, Si,…) in higher amount. Moreover, there are different approaches being investigated, which are based on the principle of inert coatings. Several deposition techniques and compositions have been proposed and tested. In this work the Physical

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Vapour Deposition (PVD) method was used and 3 different compositions were used, TiN, FeAl, FeCrAl. The TiN should act as an inert material that is meant not to react with Lead, on the other side FeAl and FeCrAl coating have the capacity of forming protective Al2O3. However, the PVD technique was not successful on depositing the Al2O3 former coatings, because of the oxidation of the Al2O3 during deposition prevented the adhesion of the coating on the substrate.In order to observe the different coatings deposited on 15-15Ti and to analyse the interface between coating and substrate, Light Optical Microscopy and Scanning Electron Microscopy were performed together with characterization of local mechanical properties by nanoindentation measurements.

804

On the effect of MOX fuel conductivity in predicting melting in FR fresh fuel by means of TRANSURANUS code

Aly Ahmed1, Davide Rozzia2, Alessandro Del Nevo3, Christophe Demaziere1

1 ChalmersUniversityofTechnology,Kemirägen4,SE-41296Goeteborg,Sweden2 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,Italy3 ENEACRBrasimone,LocalitaBrasimone,40032Camugnano(BO),[email protected]

The capability of the fuel to operate at high power without melting is important to Fast Reactor’s since reactor design limits normally require that there be a low probability of fuel melting during steady-state operation, including overpower conditions. This requirement has a direct effect on the steady-state power limit of the fuel pin and hence on the reactor power. The development of computational tools that are able to capture the occurrence of high temperature phenomena and mechanisms is thus an important step in reducing the margin of conservatisms increasing the reactor efficiency. Among the experiments that were conducted for this purpose, HEDL P-19 experiment has been selected and simulated using TRANSURANUS code to exploit its capability to capture the inception of MOX fuel melting. The experiment included 8 fresh pins with cladding outside diameters 5.84 mm, and 8 fresh 6.35 mm OD pins. It was performed during 1971 to investigate the effect of the initial fuel-to-cladding diametric gap size (from 0.86 to 0.254 mm) on the linear heat rate needed to initiate incipient melting at beginning-of-life. All 5.84 mm OD pins with fuel-to-cladding gaps equal to or less than 0.14 mm had no fuel melting. The remaining 5.84 mm OD pins and all the 6.35 mm OD pins experienced partial fuel melting.This activity consists of two parts and is limited to two representative fuel rods. The main objective of the first part is to assess the capability of TRANSURANUS to predict the measured melting heights of the tested rods and its implications on the thermal conductivity correlations implemented in the code. The second part included modifications that targeted the high temperature thermal conductivity term in two TRANSURANUS correlations. The modifications were incorporated into the code that was recompiled.

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805

Porosimetry of ZrO2 Scales Formed During Oxidation of Zr-Based Fuel Claddings in Nuclear Severe Accident ConditionsFlorian Haurais1, Emilie Beuzet1, Martin Steinbrück2, Yunxiao Wu2, Antoine Ambard1, Eric Simoni3, Mohamed Torkhani1

1 EDF-R&D,1,avenueduGeneraldeGaulle,92141ClamartCedex,France2 KarlsruheInstituteofTechnology,P.O.Box3640,76021Karlsruhe,Germany3 InstitutdePhysiqueNucléaireUniversitéParis,SudXI,F-91406Orsay,F-91406Orsay,[email protected]

In Nuclear Power Plants (NPP), a Severe Accident (SA) is a hypothetical sequence of events that can lead to reactor core degradation and to possible radiotoxic releases in the environment. If a breach of the primary circuit induces a loss of coolant, the residual power may successively lead to temperature rises, water vaporization, core uncovery, and oxidation by steam of core components and materials. In particular, the oxidation of the numerous fuel claddings, made of zirconium (Zr) alloys and constituting a containment barrier in NPP, is a key process which influences the whole SA progression.At high temperatures typical of SA, the oxidation kinetics of these Zr-based claddings under steam is generally parabolic due to the progressive thickening of a protective oxide (ZrO2) scale. Moreover, the Zr oxidation involves two phenomena inside this growing oxide layer: formation of closed pores at grain boundaries and accumulation of stresses due to a high Pilling-Bedworth ratio. Hence, at specific temperatures, a ZrO2 allotropic phase change eases the cracking of this oxide scale which may become porous and not protective anymore. In such conditions, the oxidation kinetics of Zr-based claddings can change from parabolic to linear or even accelerated.Additionally, the temperature increase can lead core materials to melt and to collapse down to the vessel bottom. Then, if a vessel failure occurs and induces air ingress into the reactor core, oxygen and nitrogen both react with Zr-based claddings, causing not only oxidation of Zr but also formation and re-oxidation of nitrides (ZrN). These coupled and self-sustained chemical reactions enhance the deterioration of oxide scales and induce a rise of their porosity.In order to quantify this porosity, a series of two-step experiments was conducted.First, Zr-based cladding samples were corroded in various conditions: at several temperatures, in some air-steam mixtures, for different durations. The main thermal effects on reaction kinetics and the high impact of air on the cladding degradation were all confirmed by experimental results.Then, porosity measurements by Hg intrusion were realized for the first time on such corroded cladding samples. In all atmospheres, it was pointed out that temperatures around 1250K lead to particularly porous oxide layers. Moreover, it was confirmed that the presence of air strongly enhances the oxide cracking: ZrO2 scales were more porous when formed in air-steam mix than under pure steam. Finally, it was shown that in all conditions, the volume percentage of the oxide porosity continuously rises during the corrosion process.

Keywords: Severe Accident, Zr-based fuel claddings, high-temperature oxidation, air ingress, oxide layer porosity, corrosion experiments, porosimetry by Hg intrusion

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806

Detailed modelling of the thermal radiation shields for applications in superconducting magnetic energy storage

systemsMihai Nicolae Anghel1, Marian Curuia2, Adrian Badea3

1 NationalR&DInstituteforCryogenicsandIsotopicTechnologies,UzineiStreet,no.4,240050-Rm. Valcea,Romania2 NationalResearchandDevelopmentInstituteforEnvironmentalProtectionEnvironmentalRadioactivity Laboratory,SplaiulIndependentei294,sector6,RO77703Bucharest,Romania3 UniversityPolitehnicaofBucharest,312,SplaiulIndependentei,Bucharest,[email protected]

Superconducting magnetic energy storage technology has the potential to bring real power storage characteristic to applications in electric grids, and specially in the future’s Smart Grids. A superconducting magnetic energy system consists of the following components: the superconducting coil, the power conditioning system, the control unit and the cryogenic cooling system. In this work are presented theoretical and experimental aspects about the heat transport by radiation with applications in superconducting magnetic energy storage systems. The objective of this paper is to show the importance of the thermal shields for reducing the influence of the radiation and also is to show the effect of the vacuum pressure on the thermal loading of the cryogenic cooling system. The numerical and experimental investigations are presented for four different thermal shields. These thermal shields were tested on the Sumitomo SRDK Series closed cycle helium cryocooler.

807

Calculation of Intergranular Stress and Strain Distributions in Neutron-Irradiated Stainless Steel Aggregate Model

Samir El Shawish, Leon CizeljJožefStefanInstitute,Jamovacesta39,1000Ljubljana,[email protected]

When austenitic stainless steel in Light Water Reactors (LWR) is subjected to neutron radiation for longer times a degradation of mechanical properties is observed, such as a drastic decrease of fracture toughness and an increased susceptibility to stress corrosion cracking. This last phenomenon is referred to as Irradiation Assisted Stress Corrosion Cracking (IASCC). The key parameters affecting both initiation and propagation of intergranular cracks in LWR environment can be divided into three groups: stress level, mechanical behavior of the material along with its chemical composition, and LWR environment. Recently, researchers from CEA, France, have developed a micromechanical crystal plasticity model to describe a nonlinear mechanical response of austenitic stainless steel subjected to neutron irradiation. The model, based on dislocation dynamics inferred mechanisms and finite strain theory, is able to capture the irradiation-induced hardening followed by softening during plastic deformation. Hardening appears to be a key factor but is not sufficient alone to explain IASCC. Change of deformation mode, from homogeneous deformation at low level of irradiation to heterogeneous and localized deformation at higher doses, has been highlighted as another key mechanism to IASCC. As argued in several studies, localized deformation promotes intergranular oxide penetration due to exposure to the primary water. Since oxide penetration weakens grain boundaries they may eventually fail at sufficiently high stress levels. It is the purpose of this study to estimate intergranular stress and strain levels in stainless steel as a function of mechanical loading

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and irradiation dose. In this respect, the recently developed crystal plasticity model from CEA, France, is implemented in Abaqus finite element solver to simulate a tensile response of the realistic stainless steel wire aggregate model composed of 377 grains. Local stress and strain concentrations calculated at grain boundaries are then used to predict IASCC initiation locations on a wire model surface.

808

Heat Exchanger Tube Cutting SystemMatej Pleterski1, Joško Valentinčič2, Izidor Sabotin2

1 NUMIPd.o.o.,Cestakrškihžrtev135E,8270KRŠKO,Slovenia2 FacultyofMechanicalEngineering,UniversityofLjubljana,Aškerčeva6,1000Ljubljana,[email protected]

Maintenance or periodic inspection of heat exchangers in nuclear power plants in some cases assumes tube sampling in order to determine the mechanism and cause of degradation. There are many obstacles and requirements due to limited access and possible radiation. At first these narrow tubes must be cut from inside out in usually great depths with harmless or zero debris left in the system. Additionally, the damage of the surrounding tubes has to be avoided. The end of the tube is usually expanded into a tube sheet, thus loosening up of that end is also often required in order to successfully extract the tube. The paper presents development, testing and optimization of the newly developed dry EDM based portable tube cutting device, as well as selection of suitable electrical generator and optimization of cutting process parameters by performing design of experiments using the Taguchi method. The apparatus consists of: an electrical pulse generator, an electro-motor, a wire rope for torque and current transmittal, a vacuum system and an unique cutting head. The head comprises specially designed bearing and electrode construction where gap between electrodes and pipe inner wall is assured without any servo controllers and is thus an object of patenting process. The presented device contributes to company’s heat exchanger services field significantly and has already proved its effectiveness and reliability in practical application, namely, it was successfully used for cutting during tube sampling of main turbine condenser tubes of Unit #1, NPP Cernavoda, Romania.

809

Thermo-mechanical model of a pipe under thermal fatigueFrancesco DolciJointResearchCentreoftheEuropeanCommission,Westerduinweg3,1755ZGPetten,[email protected]

Thermal and mechanical loads of components are crucial factors which have to be taken into consideration during the design and operation of a Nuclear Power Plant (NPP).A number of piping components of a NPP can be subjected to a thermal down-shock. In such cases the surface of a component is subjected to strong tensile stresses, while the rest of the material self-equilibrates, thus causing a distribution of stresses across the wall thickness. The nature of the material used, the specimen's geometry and the thermal-shock conditions, all have an impact on the levels of stress and consequently the integrity of the component.The JRC-IET in Petten has developed a facility to test the impact of realistic thermal fatigue loads on damage initialization and evolution in pipe components. In the test facility a steel pipe is heated by an external heating coil while cooling water is cyclically injected into the pipe. After a number of loading cycles surface cracks initialize on the inner surface and start to propagate through the thickness.Finite Element modelling of the tested specimen is run in parallel and used to better assess experimental

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data. A model allows for a wider parametric analysis without relying on expensive and time consuming tests. However, the model needs to be validated first. This also includes the proper definition of loading/boundary conditions and the material properties. The temperature and thermal properties have an obvious impact on a thermal-shock analysis.The current paper presents a combined computational fluid dynamics (CFD) and structural mechanical thermal fatigue model. Heat transfer coefficients and sink temperatures are obtained with a CFD and then imposed as boundary conditions on the structural mechanical model. The results are then checked against experimental temperature profiles. Comparison between the results of a P91 ferritic-martensitic alloy steel specimen and 316L austenitic stainless steel specimen is done.

810

Vortex Robot for Rapid Low Cost Scanning and Improved Non-Destructive Testing of Large Concrete Structures

Petar Mateljak1, Estefania Artigao2, Eleni Cheilakou3, Vassilis Kappatos2, Alvaro Garcia4, Marko Budimir1

1 INETEC-InstituteforNuclearTechnology,Dolenica28,10250Zagreb,Croatia2 BrunelInnovationCentre,BrunelUniversity,London,UnitedKingdom3 UniversityofIoannina,Greece4 InnoTecUKLtd,[email protected]

Large public infrastructure facilities around the world like dams, cooling towers and bridges use cement as the main building material. Although made from a durable material many of these large assets have begun to age and are in need of periodic inspection to ensure their integrity. Current inspection routine involves setting up scaffolds that have to be moved around for personnel to reach to the whole of the surface. The VORTEXSCAN project has developed a vortex robot that is used to autonomously navigate the vertical surfaces of such structures while deploying a combination of non-destructive inspection techniques. Air suction through a nozzle of a specific geometry creates a vortex and initiates a force that attaches the robot to a vertical surface while its wheels move it around. Novel phased array Ground Penetrating Radar (GPR) technique is one of the two NDT equipment carried by the robot. Low frequency ultrasonic technique is also developed and an innovative fusion of data from the two techniques is performed. The data fusion is used to diminish the drawbacks of each technique and create an accurate representation of the underlying material and its defects. VORTEXSCAN project goal is to create a system that can rapidly and economically inspect large vertical concrete structures saving huge amounts currently spent during the manual periodic inspections. The system will be used on EUs electrical power industry structures such nuclear reactor containment, dams and cooling towers as a large number of these infrastructure types are beginning to age; some of them having been built nearly a century ago.

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811

Modeling and Assessment of PCI in LWR FuelDavide Rozzia1, Alessandro Del Nevo2, Lelio Luzzi3

1 UniversitadegliStudidiPisa,DipartimentodiIngegneriaMeccanicaNucleareedellaProduzione, LargoLucioLazzerino1,56100Pisa,Italy2 ENEACRBrasimone,LocalitaBrasimone,40032Camugnano(BO),Italy3 PolitecnicodiMilano-departmentofenergy,ViaLaMasa34,20156Milano,[email protected]

Investigations on fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. In this framework, OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments database”. This database includes: Super-Ramp Project (BWR and PWR), BWR Inter-Ramp Project and CNEA MOX Experiment. The common objectives of these experimental programmes are to investigate the Pellet Cladding Interaction (PCI) failure mechanism and associated phenomena during power ramp tests with the aim to establish LWR fuel failure-safe operating limits. The fuel type comprises PWR, BWR and MOX designs. The burn-up ranges from zero to 44 MWd/kgU. The analyses are carried out in the framework of the IAEA-CRP FUMEX-III.The objective of the activity is to provide an overview of the TRANSURANUS code capabilities in simulating the PCI phenomenon with particular reference to the predictability of the cladding failure. The analysis summarizes the main results achieved after the simulations of the above mentioned databases that include a representative amount of fuel rods (74) of various types and design. The different irradiation histories and burn-ups are accounted for. Focus is given on the main variables, which are involved or may influence the cladding failure and characterize the power ramp tests.

812

Effect of tensile strain on micro-structure of irradiated core internal material

Hygreeva Kiran NamburiResearchcentreRez,Hlavni130,25068Husinec-Řež,[email protected]

Irradiation Assisted Stress Corrosion Cracking [IASCC] is one of the most significant environmental degradation in the internal components made from Austenitic stainless steel. This mechanism is still not fully understood and there are no suitable criteria for prediction of the damage during operation.In this work, core basket material 08Ch18N10T austenitic stainless steel acquired from decommissioned NPP Nord / Greifswald Unit 1, VVER 440-230 type, operated for 15 years and irradiated at 5.2 dpa is studied. This material was tensile tested at two different test temperatures and strain rates in air and at the elevated temperature under the water environment. SEM observations of the fracture surface documented ductile fracture of the samples tested in air, but areas of IASCC tested in water.This paper emphases on the microscopic examination results from the mechanically tested samples to determine the underlying IASCC physical damage process. TEM observations of thin foils made from the gauge sections that are closer to the fractured surface of the specimen aimed to find variances in interaction of dislocations and grain boundaries owing to different test conditions.

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Kot sistemski operater slovenskega elektroenergetskega prenosnega omrežja smo strokovnjaki za prenos električne energije. Ljudje z znanjem in izkušnjami, ki skrbimo za njen varen, zanesljiv in neprekinjen prenos. Strateško, odgovorno in trajnostno načrtujemo, gradimo in vzdržujemo prenosno omrežje Republike Slovenije. Tu smo 24 ur na dan. Za električno energijo na dosegu vaše roke.

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