European LEad-Cooled TRAining reactor: structural materials and design issues
|
|
- Derick Wade
- 6 years ago
- Views:
Transcription
1 Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials JUNE 2013 IAEA HQ, VIENNA, AUSTRIA European LEad-Cooled TRAining reactor: structural materials and design issues S. Bortot, J. Ejenstam, M. Pukari, E. Suvdantsetseg, P. Szakalos, J. Wallenius Kungliga Tekniska Högskolan!
2 Outline! ELECTRA: European Lead-Cooled TRAining reactor! Why?! Where?! How?! Choice of materials! Fuel! Cladding, components and structures! Core design and modeling issues! Concluding remarks
3 ELECTRA: European Lead-Cooled TRAining reactor WHY?! Education and training facility! Test bed for LFR technology! Research on fast reactor dynamics! R&D on fuel recycle & manufacture
4 ELECTRA: European Lead-Cooled TRAining reactor WHERE?! ELECTRA-FCC to be built in Oskarshamn CLAB! ELECTRA: 0.5 MW LFR with (Pu,Zr)N fuel! OKG power plant premises OKG! Fuel cycle facilities for ELECTRA! Premises of CLAB interim storage CLAB
5 ELECTRA: European Lead-Cooled TRAining reactor HOW?! 0.5 MW LFR! 100 % natural convection! (Pu,Zr)N fuel! Core size 30 x 30 cm! Reactor vessel approximately 1 x 3 m! Tentative cost: 52 M, including FFC
6 Choice of materials FUEL - general! Only U free inert matrix fuels can provide the small core size necessary to achieve full heat removal by natural convection in an LFR! Several inetr matrix fuels exhibit good performance under irradiation:! PuO2-MgO (BOR-60)! (Pu,Am)O2-Mo (HFR)! (Pu,Zr,Y)O2 (Halden)! (Pu,Zr)N (BOR-60, JMTR, HFR, ATR)! Inert matrix nitride fuels offer the highest Pu density, hence small core size for ELECTRA
7 Choice of materials FUEL - properties Thermal conductivity [W/m/k]! Thermal conductivity for (Pu 0.4,Zr 0.6 )N measured by VNIINM (Pu 0.4,Zr 0.6 )N T [ C] ! High temperature stability test carried out under vacuum, argon and nitrogen Weight loss (Pu 0.4,Zr 0.6 )N ZrN 2300 C 26.4% 1.9% 2300 C 2.7% 1.3% 2200 C 1.7% 0.9%
8 Choice of materials FUEL manufacturing process! ZrN to be fabricated from metallic zirconium (to reduce impurities)! PuN to be fabricated from PuO2 (availability)! Mixing and heat treatment à solid solution! Process optimization carried out in collaboration with JAEA (Pu,Zr)N with 88% density! Less than 0.3 wt. % oxygen & carbon impurities achievable by using standard equipment! 2 fuel pins for test irradiation & fuel qualification to be fabricated at PSI
9 Choice of materials FUEL irradiation performance: CONFIRM! (Pu 0.3,Zr 0.7 )N fuel fabricated by PSI within CONFIRM; oxide source material! 20 % initial porosity CONFIRM! Irradiation to 10 % burnup in HFR! Linear rating: kw/m! Gas release: < 5 % Xenon, 80 % Helium! Swelling: 0.9 % per percent Pu burnup! No internal corrosion
10 Choice of materials FUEL irradiation performance: BORA-BORA! Two-phase PuN-ZrN fuel fabricated by VNIINM! Metallic source material! 16 % initial porosity BORA-BORA! Irradiation in BOR-60 up to 19.4 % burnup! Linear rating: < 20 kw/m! Gas release: < 1 %! Swelling: < 0.1 % per percent Pu burnup! Internal corrosion: 15 microns; oxide phase observed in PIE
11 Choice of materials FUEL fabrication! 880 kg separated Pu, owned by OKG, in Sellafield! A Pu fuel fabrication lab is under commissioning in Chalmers! Capacity: one fuel pellet per day! In 2012 a Pu conversion facility was licensed, commissioned and operated in Studsvik! Capacity: 200 g Pu per day = one fuel pin for ELECTRA! Relicensing for ELECTRA fuel fabrication under discussion
12 Choice of materials CLADDING general 400 Stress [MPa]! Cladding creep rupture is a safety limiting design parameter in ELECTRA T91 12R72 (15/15Ti)! Use of ELECTRA for transient tests requires high creep rupture strength! Austenitic steel selected 0 LMP [10 3 ] ! Sandvik developed the 12R72 grade for Phénix in early 70 s Cr Ni Mn Mo Ti Si ! Fabrication resumed for MYRRHA! Extensive set of creep data exists ( hr)
13 Choice of materials CLADDING swelling: dose rate dependence! EoL dose for ELECTRA cladding tubes < 40 dpa! Swelling data for cold worked 15-15Ti indicate swelling threshold > 100 dpa Swelling of 50 dpa [%] 5x10-7 dpa/s! Dose rate dependence of swelling for austenitics is significant! Swelling larger at same dose for lower dose rates (Budylkin 2004) 10 16x10-7 dpa/s 5 Si fraction [wt %] ! ELECTRA dose rate: 1 x 10-7 dpa/s! Swelling threshold to be confirmed
14 Choice of materials CLADDING corrosion! Max cladding temperature under nominal operation: 830 K (540 C) GESA-FeCrAlY Austenitic steel GESA-FeCrAlY + T91! Alumina (or silica) protection necessary for operation beyond 8000 hours! Al-bearing bulk steel has low creep resistance! GESA protection of austenitic steels is applicable today for short cladding tubes if limited number is to be produced! GESA was tested on 15-15Ti in LBE at 600 C for up to 8000 hours! ELECTRA cladding: 50 cm x 400 pieces
15 Choice of materials HEAT EXCHANGERS! F/M steels preferred due to lower thermal expansion! Bulk Fe10CrAl-RE under development in collaboration with Sandvik! hours tests of Fe10CrXAl-RE in lead performed at 550 C & 10-7 wt.% Oxigen 200 nm Fe10Cr6Al-RE C! 6 wt.% Al ensures thin (0.1 micron), stable and protective alumina scale formation! Al content under optimization (welding issues)! Thermal ageing tests are carried out to investigate potential for embrittlement
16 Core design and modeling issues CORE CONFIGURATION! (Pu 0.4,Zr 0.6 )N fuel, ~ 70 kg Pu from spent UOX! 397 fuel pins, D clad = 12.6 mm! Active core dimensions: ~ 30 x 30 cm! Shutdown and reactivity compensation using 12 rotating B 4 C drums.! Linear rating: ~4 kw/m! Core life: 14 full power years 10 B4C/steel drums! Burnup: ~ 5% fission in actinides! Maximum dose: dpa
17 Core design and modeling issues CONTROL SYSTEM DESIGN! Single drum rotation should not lead to cladding creep rupture! SAS4A transient over-power simulation indicates limit of reactivity insertion at 1.7$ (450 pcm), for a rate of 1$/s.! Total reactivity loss compensation: > 5000 pcm! 12 drums necessary! Similar to Russian space reactor concept (TOPAZ/Yenisey)
18 Core design and modeling issues SAFETY PARAMETERS Parameter BOL EOL βeff 270 pcm 240 pcm KDoppler ~ 0 ~ 0 αpb (core) -0.4 pcm/k -0.4 pcm/k αpb (global) -1.8 pcm/k -2.5 pcm/k Pb void (core) pcm pcm αaxial -0.4 pcm/k -0.4 pcm/k αradial -1.5 pcm/k -1.3 pcm/k! Very hard neutron spectrum! Zero Doppler feedback! Large leakage! Negative coolant temperature coefficient! Large negative axial temperature feedback! Coolant temperature coefficient more negative at EOL, in spite of 8% Am in the fuel
19 Core design and modeling issues CHALLENGING IN MODELING: REPRODUCTION TIME! Effective neutron reproduction time affected by presence of large, nonabsorbing reflector (lead)! 67% of all neutrons leak into the reflector! 26% of leaked neutrons return to core!! Deterministic calculations face problems related to convergence of (adjoint) flux in reflector! Cross sections for adjoint Monte Carlo not available at KTH! Two options:! Time dependent Monte Carlo (MCNP5) à Λ = 61±2 ns (dynamic repr. time)! Perturbation with 1/v absorber (Serpent) à Λ = 80±2 ns (static repr. time)
20 Core design and modeling issues CHALLENGING IN MODELING: AXIAL EXPANSION IN SAS4A/SASSYS-1! SAS4A permits to model axial dependence of axial expansion worth, using an axial distribution of fuel worth! Change in reactivity = sum of changes in fuel density x fuel worth! Increase of core height not taken into account! In small cores, a significant error arises! ELECTRA:! density = -2% -> k = pcm! H = +2% -> k = pcm
21 Core design and modeling issues CHALLENGING IN MODELING: NATURAL CIRCULATION! BELLA benchmarked towards SAS4A for 0.1$ reactivity insertion! Maximum power and temperatures agree reasonably well.! Temperatures decrease faster in BELLA! Significant discrepancy in natural convection flow rate
22 Concluding remarks! A small LFR may be designed to operate on 100% natural convection! Sandvik s 12R72 grade of 15-15Ti delivers high creep resistance! GESA treatment potentially ensures adequate corrosion protection! Bulk Fe10CrXAl-RE considered for heat exchangers! Open issues:! Dose rate dependence of swelling in austenitic cladding materials! Al-limit for weldability of FeCrAl-RE steels! Modeling of ELECTRA transients reveals major challenges:! Neutron reproduction time! Axial expansion coefficient! Natural convection flow rates
23 Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials JUNE 2013 IAEA HQ, VIENNA, AUSTRIA Thank you for your kind attention!
Simulation of large and small fast reactors with SERPENT
Simulation of large and small fast reactors with SERPENT Janne Wallenius, Erdenechimeg Suvdantsetseg, Sara Bortot Milan Tesinsky & Youpeng Zhang Reactor Physics Kungliga Tekniska Högskolan R&D activities
More informationELECTRA-FCC: An R&D centre for Generation IV systems in Sweden. Janne Wallenius Professor Reactor Physics, KTH
ELECTRA-FCC: An R&D centre for Generation IV systems in Sweden Janne Wallenius Professor Reactor Physics, KTH What do Generation IV nuclear systems offer? Recycle of U-238 from spent fuel and enrichment
More informationTransmutation. Janne Wallenius Professor Reactor Physics, KTH. ACSEPT workshop, Lisbon
Transmutation Janne Wallenius Professor Reactor Physics, KTH Why would one want to transmute high level nuclear waste? Partitioning and transmutation of Pu, Am & Cm reduces the radio-toxic inventory of
More informationSEALER: A small lead-cooled reactor for power production in the Canadian Arctic
1 IAEA-CN245-431 SEALER: A small lead-cooled reactor for power production in the Canadian Arctic J. Wallenius 1, S. Qvist 1, I. Mickus 1, S. Bortot 1, J. Ejenstam 1,2, P. Szakalos 1 1 LeadCold Reactors,
More informationEP-450 Steel as Cladding Material for Fast Neutron Reactor Fuel Rods
EP-450 Steel as Cladding Material for Fast Neutron Reactor Fuel Rods A.Povstyanko, V.Prokhorov, A. Fedoseyev, F.Kryukov JSC SSC RIAR Patent 677544 15.11.77 Material C Si Mn Composition, % mass. Cr Ni Mo
More informationSTRENGTH OF STEELS EXPOSED TO HEAVY LIQUID METALS
IAEA INPRO COOL project Activity 6: Components for service in intimate contact with high temperature coolants (LM and MS) 6.1 Experimental study on components and various materials that are in contact
More informationAEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th )
AEN WPRS Sodium Fast Reactor Core Definitions (version 1.2 September 19 th ) David BLANCHET, Laurent BUIRON, Nicolas STAUFF CEA Cadarache Email: laurent.buiron@cea.fr 1. Introduction and main objectives
More informationIrradiation of Metallic Fuels With Rare Earth Additions for Actinide Transmutation in the ATR Experiment Description for AFC-2A and AFC-2B
INL/EXT-06-11707 Irradiation of Metallic Fuels With Rare Earth Additions for Actinide Transmutation in the ATR Experiment Description for AFC-2A and AFC-2B S.L. Hayes T.A. Hyde W.J. Carmack November 2006
More informationLEU Conversion of the University of Wisconsin Nuclear Reactor
LEU Conversion of the University of Wisconsin Nuclear Reactor Paul Wilson U. Wisconsin-Madison Russian-American Symposium on the Conversion of Research Reactors to Low Enriched Uranium Fuel 8-10 June 2011
More informationPre-Conceptual Core Design of a LBE-Cooled Fast Reactor (BLESS) Ziguan Wang, Luyu Zhang, Eing Yee Yeoh, Linsen Li, Feng Shen
Pre-Conceptual Core Design of a LBE-Cooled Fast Reactor (BLESS) Ziguan Wang, Luyu Zhang, Eing Yee Yeoh, Linsen Li, Feng Shen State Power Investment Corporation Research Institute, Beijing 102209, P. R.
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
NEUTRONIC ANALYSIS OF THE CANDIDATE MULTI-LAYER CLADDING MATERIALS WITH ENHANCED ACCIDENT TOLERANCE FOR VVER REACTORS Ondřej Novák 1, Martin Ševeček 1,2 1 Department of Nuclear Reactors, Faculty of Nuclear
More informationAn Introduction to the Engineering of Fast Nuclear Reactors
An Introduction to the Engineering of Fast Nuclear Reactors This book is a resource for both graduate-level engineering students and practicing nuclear engineers who want to expand their knowledge of fast
More informationIrradiation Testing of Structural Materials in Fast Breeder Test Reactor
Irradiation Testing of Structural Materials in Fast Breeder Test Reactor IAEA Technical Meet (TM 34779) Nov 17-21, 2008 IAEA, Vienna S.Murugan, V. Karthik, K.A.Gopal, N.G. Muralidharan, S. Venugopal, K.V.
More informationFuels for advanced sodium cooled fast reactors in Russia: state of-art and prospects
Fuels for advanced sodium cooled fast reactors in Russia: state of-art and prospects L. M. Zabudko, V.M. Poplavsky (1), I.A. Shkaboura, M.V. Skupov (2), F.V. Bychkov, V.A. Kisly, F.N. Kryukov (3), B.A.
More informationMYRRHA FUEL TRANSIENT TESTS PROJECT AT THE TRIGA-ACPR REACTOR
MYRRHA FUEL TRANSIENT TESTS PROJECT AT THE TRIGA-ACPR REACTOR CSABA ROTH, BRIAN BOER*, MIREA MLADIN, ADRIAN DATCU, GEORGIANA BUDRIMAN, CALIN TRUTA Institute for Nuclear Research Pitesti, Romania * SCK
More informationFuel for sodium cooled fast reactors in Russia
Fuel for sodium cooled fast reactors in Russia L. M. Zabudko, V.M. Poplavsky (1), I.A. Shkaboura, M.V. Skupov (2), V.A. Kisly, F.N. Kryukov (3), State Corporation for Atomic Energy ROSATOM (1)Institute
More informationMCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR
U.P.B. Sci. Bull., Series D, Vol. 74, Iss. 1, 2012 ISSN 1454-2358 MCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR Mirea MLADIN 1, Daniela MLADIN 21 The paper describes the
More informationGas-cooled Fast Reactor Status and program. Pascal ANZIEU Commissariat à l énergie atomique Atomic Energy Commission France
Gas-cooled Fast Reactor Status and program Pascal ANZIEU Commissariat à l énergie atomique Atomic Energy Commission France Nuclear Energy Division P. Anzieu - GFR Status 1 GFR: an alternative Fast Neutrons
More informationThe new material irradiation infrastructure at the BR2 reactor. Copyright 2017 SCK CEN
The new material irradiation infrastructure at the BR2 reactor The new material irradiation infrastructure at the BR2 reactor Steven Van Dyck, Patrice Jacquet svdyck@sckcen.be Characteristics of the BR2
More informationNeutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar
Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar Md. Quamrul HUDA Energy Institute Atomic Energy Research Establishment Bangladesh Atomic Energy
More informationDesign and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 32[9], pp. 834-845 (September 1995). Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
More informationEvolution of Nuclear Energy Systems
ALLEGRO Project 2 Evolution of Nuclear Energy Systems 3 General objectives Gas cooled fast reactors (GFR) represent one of the three European candidate fast reactor types. Allegro Gas Fast Reactor (GFR)
More informationInterfaces: Corrosion in Pb-alloy cooled nuclear reactors and advanced mitigation measures
Interfaces: Corrosion in Pb-alloy cooled nuclear reactors and advanced mitigation measures and G. Müller KIT KIT Universität des Landes Baden-Württemberg und nationales Forschungszentrum in der Helmholtz-Gemeinschaft
More informationFast Reactor Operating Experience in the U.S.
Fast Reactor Operating Experience in the U.S. Harold F. McFarlane Deputy Associate Laboratory Director for Nuclear Science and Technology www.inl.gov 3 March 2010 [insert optional photo(s) here] Thanks
More informationOn going issues on structural
On going issues on structural materials of LFR Luigi Debarberis Institute for Energy (IE) Petten, The Netherlands http://www.jrc.ec.europa.eu Trieste, April 2009 1 On going issues on structural materials
More informationAnalysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor
Analysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor G. Bandini (ENEA/Bologna) E. Bubelis, M. Schikorr (KIT/Karlsruhe) A. Alemberti, L. Mansani (Ansaldo Nucleare/Genova) Consultants Meeting:
More informationNumerical Benchmark Results for 1000MWth Sodium-cooled Fast Reactor
Numerical Benchmark Results for 1000MWth Sodium-cooled Fast Reactor T. K. Kim and T. A. Taiwo Argonne National Laboratory February 13, 2012 Second Meeting of SFR Benchmark Task Force of Working Party on
More informationOverview, Irradiation Test and Mechanical Property Test
IAE R&D Program Progress Report Development Project of Supercritical-water Cooled Power Reactors Overview, Irradiation Test and Mechanical Property Test Shigeki Kasahara Hitachi, Ltd. Toshiba Corp. Hokkaido
More informationnuclear science and technology
EUROPEAN COMMISSION nuclear science and technology Co-ordination and Synthesis of the European Project of Development of HTR Technology (HTR-C) Contract No: FIKI-CT-2000-20269 (Duration: November 2000
More informationLFR safety features. through intrinsic negative reactivity feedbacks
LFR safety features through intrinsic negative reactivity feedbacks Giacomo Grasso UTFISSM Technical Unit for Reactor Safety and Fuel Cycle Methods Leader of Core Design Work Package in the EURATOM FP7
More informationSERPENT activities at VUJE, a.s.
SERPENT activities at VUJE, a.s. Tomáš Chrebet Amine Bouhaddane František Čajko Michal Sečanský Radoslav Zajac 26-29 September 2016 6th International Serpent User Group Meeting at Milan SERPENT activities
More informationSurface and Corrosion Science, Kungliga Tekniska Högskolan, Stockholm, Sweden. Reactor Technology, Kungliga Tekniska Högskolan, Stockholm, Sweden
ELECTRA-FCC: A Swedish R&D centre for Generation IV systems Janne Wallenius a, Erdenechimeg Suvdantsetseg a, Sara Bortot a, Mikael Jolkkonen a, Merja Pukari a, Jesper Ejenstam b, Peter Szakalos b, Roman
More informationAnalysis of Core Physics Test Data and Sodium Void Reactivity Worth Calculation for MONJU Core with ARCADIAN-FBR Computer Code System
FR09 - International Conference on Fast Reactors and Related Fuel Cycles Analysis of Core Physics Test Data and Sodium Void Reactivity Worth Calculation for MONJU Core with ARCADIAN-FBR Computer Code System
More informationTrends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors
Trends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors The Tenth OECD Nuclear Energy Agency Information Exchange Meeting on Actinide and Fission
More informationReactivity requirements can be broken down into several areas:
Reactivity Control (1) Reactivity Requirements Reactivity requirements can be broken down into several areas: (A) Sufficient initial reactivity should be installed to offset the depletion of U 235 and
More informationDevelopment of Low Activation Structural Materials
Materials Challenge for Clean Nuclear Fusion Energy Development of Low Activation Structural Materials T. Muroga National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan Symposium on Materials
More informationIrradiation Facilities at the Advanced Test Reactor International Topical Meeting on Research Reactor Fuel Management Lyon, France
Irradiation Facilities at the Advanced Test Reactor International Topical Meeting on Research Reactor Fuel Management Lyon, France S. Blaine Grover Idaho National Laboratory March 12, 2007 Agenda Advanced
More informationCore Modification for the High Burn-up to Improve Irradiation Efficiency of the Experimental Fast Reactor Joyo
Core Modification for the High Burn-up to Improve Irradiation Efficiency of the Experimental Fast Reactor Joyo S. Maeda, M. Yamamoto, T. Soga, T. Furukawa and T. Aoyama Oarai R&D Center Japan Atomic Energy
More informationTask 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS
Task 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS Volkan Seker, Matt Neuman, Nicholas Kucinski, Hunter Smith, Tom Downar University of Michigan May 24,
More informationStructural materials for Fusion and Generation IV Fission Reactors
Hungarian Academy of Sciences KFKI Atomic Energy Research Institute Structural materials for Fusion and Generation IV Fission Reactors Ákos Horváth Materials Department akos.horvath@aeki.kfki.hu EFNUDAT
More informationJournal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.
Burn-up credit in criticality safety of PWR spent fuel Rowayda F. Mahmoud 1, Mohamed K.Shaat 2, M. E. Nagy 3, S. A. Agamy 3 and Adel A. Abdelrahman 1 1 Metallurgy Department, Nuclear Research Center, Atomic
More informationStainless Steel 310/310S (UNS S31000/ UNS S31008)
Stainless Steel 310/310S (UNS S31000/ UNS S31008) Austenitic Stainless Steel 310/310S offers excellent resistance to oxidation up to 2000oF. It is a low grade steel that prevents embrittlement and sensitization.
More informationSafety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
FR13 - TECHNICAL SESSION 3.5: Fast reactor safety: post-fukushima lessons and goals for next-generation reactors Paper n. IAEA-CN-199/260 Safety Analysis Results of Representative DEC Accidental Transients
More informationCONQUER CORROSION. Key issues of the lead-cooled fast reactor design VATTENFALL RESEARCH AND DEVELOPMENT AB
CONQUER CORROSION Key issues of the lead-cooled fast reactor design Mathias Hareland, MSc Thesis Uppsala University VATTENFALL RESEARCH AND DEVELOPMENT AB Report Number U 10:124 2011-03-15 Distribution
More information1 IAEA-CN The primary author's address: 1 Introduction
1 IAEA-CN245-062 Development of innovative fast reactor nitride fuel in Russian Federation: state-of-art Grachev A.F. 1), Zabudko L.M. 1) Zvir E.A. 2), Zozulya D.V. 3), Ivanov Yu.A. 4), Kryukov F.N. 2)
More informationWestinghouse Advanced Doped Pellet Characteristics and Irradiation behaviour
Westinghouse Advanced Doped Pellet Characteristics and Irradiation behaviour Karin Backman 1, Lars Hallstadius 1 and Gunnar Rönnberg 2 1. Westinghouse Electric Sweden, 2. OKG AB Sweden IAEA - Technical
More informationFast Neutron Reactors & Sustainable Development
International Conference on Fast Reactors and Related Fuel Cycles December 7th 11th, 2009, Kyoto, Japan Fast Neutron Reactors & Sustainable Development Jacques BOUCHARD Chairman of the Generation IV International
More informationAN INVESTIGATION OF TRU RECYCLING WITH VARIOUS NEUTRON SPECTRUMS
AN INVESTIGATION OF TRU RECYCLING WITH VARIOUS NEUTRON SPECTRUMS Yong-Nam Kim, Hong-Chul Kim, Chi-Young Han and Jong-Kyung Kim Hanyang University, South Korea Won-Seok Park Korea Atomic Energy Research
More informationInnovations and Safety Ensuring in WWERs on the Base of Collaboration on the National and International Levels
IAEA INPRO DF9, Vienna 21 November 2014 Innovations and Safety Ensuring in WWERs on the Base of Collaboration on the National and International Levels Grigory Ponomarenko OKB GIDROPRESS Podolsk, Russian
More informationTechnical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors
Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors Spiral-Tube Steam Generators for compact integrated reactors. The ELSY project Vienna, December 21-22, 2011
More informationSeminar on Coolants for Fast Neutron Reactors. Frank Carré & Yves Bamberger CEA, Nuclear Energy Division, France
Seminar on Coolants for Fast Neutron Reactors Frank Carré & Yves Bamberger CEA, Nuclear Energy Division, France Maturity of Fast Reactor Types Sodium fast reactor ~400 reactor.year of operating experience
More informationTools and applications for core design and shielding in fast reactors
Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials, June 12-14, 2013 Tools and applications for core design and shielding in fast reactors Presented by: Reuven Rachamin
More informationStatus of Compatibility Facilities and Experiments on LBE with CLEAR-I Structural Materials
Status of Compatibility Facilities and Experiments on LBE with CLEAR-I Structural Materials Presented By Zhizhong JIANG Contributed by FDS Team Institute of Nuclear Energy Safety Technology Chinese Academy
More informationFusion structural material development in view of DEMO design requirement
3 rd IAEA DEMO programme workshop 11 th 14 th May, 2015, Hefei, China Fusion structural material development in view of DEMO design requirement A case study on a RAFM steel F82H development in view of
More informationA HELIUM COOLED PARTICLE FUEL REACTOR FOR FUEL SUSTAINABILITY. T D Newton, P J Smith and Y Askan SERCO Assurance, Winfrith, Dorset, England * Abstract
A HELIUM COOLED PARTICLE FUEL REACTOR FOR FUEL SUSTAINABILITY T D Newton, P J Smith and Y Askan SERCO Assurance, Winfrith, Dorset, England * Abstract Sustainability is a key goal for future reactor systems.
More informationSpecification for Phase IID Benchmark. A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA)
Specification for Phase IID Benchmark PWR-UO 2 Assembly: Study of control rods effects on spent fuel composition A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA) 1. Introduction
More informationA study of reactivity control by metallic hydrides for Accelerator Driven System
1 ADS/P4-15 A study of reactivity control by metallic hydrides for Accelerator Driven System K. Abe 1, T. Iwasaki 1, Y. Tanigawa 1 1 Tohoku University, JAPAN Email contact of main author: kazuaki@neutron.qse.tohoku.ac.jp
More informationGerman SFR Research and European Sodium Fast Reactor Project
German SFR Research and European Sodium Fast Reactor Project B. Merk Institute of Resource Ecology (FWO) Helmholtz-Zentrum Dresden-Rossendorf with special thanks to E. Fridman, G. Gerbeth (FWD), A. Vasile
More informationFull MOX Core Design in ABWR
GENES4/ANP3, Sep. -9, 3, Kyoto, JAPAN Paper 8 Full MOX Core Design in ABWR Toshiteru Ihara *, Takaaki Mochida, Sadayuki Izutsu 3 and Shingo Fujimaki 3 Nuclear Power Department, Electric Power Development
More informationComparison of characteristics of fast and thermal reactors, Role of fast reactors in Indian Nuclear Programme
Comparison of characteristics of fast and thermal reactors, Role of fast reactors in Indian Nuclear Programme K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University Joint Initiative
More informationMixed-oxide (MOX) fuel performance benchmarks
Mixed-oxide (MOX) fuel performance benchmarks L. J. Ott a,*, Terje Tverberg b, Enrico Sartori c a Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A. b OECD Halden Reactor Project, Halden, Norway
More informationChapter VI Core physics
Controlling the core of a sodium-cooled fast reactor is very simple compared to pressurized water reactors. The power, due to thermal feedback coefficients, is stable for any given position of the control
More informationNeutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor
Neutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor David Chandler and R. J. Ellis chandlerd@ornl.gov Oak Ridge National
More informationABSTRACT. 1. Introduction
Improvements in the Determination of Reactivity Coefficients of PARR-1 Reactor R. Khan 1*, Muhammad Rizwan Ali 1, F. Qayyum 1, T. Stummer 2 1. DNE, Pakistan Institute of Engineering and Applied Sciences
More informationLFR core design. for prevention & mitigation of severe accidents
LFR core design for prevention & mitigation of severe accidents Giacomo Grasso UTFISSM Technical Unit for Reactor Safety and Fuel Cycle Methods Coordinator of Core Design Work Package in the EURATOM FP7
More informationValidation of the Monte Carlo Code MVP on the First Criticality of Indonesian Multipurpose Reactor
Validation of the Monte Carlo Code MVP on the First Criticality of Indonesian Multipurpose Reactor T.M. Sembiring, S. Pinem, Setiyanto Center for Reactor Technology and Nuclear Safety,PTRKN-BATAN, Serpong,
More informationTopic 1: Fuel Fabrication. Daniel Mathers and Richard Stainsby
Topic 1: Fuel Fabrication Daniel Mathers and Richard Stainsby CEIDEN NNL meeting, Sellapark, 1 st February 2016 Level of Benefit / Ambition UK Fuel Ambition: Development of Fuels with Enhanced Safety,
More informationThe Challenge of Nuclear Reactor Structural Materials for Generation IV Nuclear Energy Systems
20th International Conference on Structural Mechanics in Reactor Technology (SMiRT 20) Espoo, Finland, August 9-14, 2009 SMiRT 20-Division 10, Paper 1586 The Challenge of Nuclear Reactor Structural Materials
More informationBurn up Analysis for Fuel Assembly Unit in a Pressurized Heavy Water CANDU Reactor
Burn up Analysis for Fuel Assembly Unit in a Pressurized Heavy Water CANDU Reactor A. A. EL-Khawlani a, Moustafa Aziz b, M. Ismail c and A. Y. Ellithi c a Physics Department, Faculty of Science, High Education,
More informationAdvanced Reactor Technology
Advanced Reactor Technology Robert N. Hill Nuclear Engineering Division Argonne National Laboratory 2012 Nanonuclear Workshop Gaithersburg, Maryland June 6, 2012 Outline Advanced Reactor Trends Small Modular
More informationImprovement of Irradiation Capability in the Experimental Fast Reactor Joyo
Improvement of Irradiation Capability in the Experimental Fast Reactor Joyo T. Soga*, T. Aoyama* and S. Soju* * Experimental Fast Reactor Department, Oarai Research and Development Center Japan Atomic
More informationE-BRITE E-BRITE. Technical Data Sheet. Stainless Steel: Superferritic GENERAL PROPERTIES PLANAR SOLID OXIDE FUEL CELLS CHEMICAL COMPOSITION
E-BRITE Stainless Steel: Superferritic (UNS 44627, ASTM Type XM-27) GENERAL PROPERTIES E-BRITE alloy is a high purity ferritic stainless steel which combines excellent resistance to corrosion and oxidation
More informationA Brief Summary of Analysis of FK-1 and FK-2 by RANNS
A Brief Summary of Analysis of FK- and by RANNS Motoe Suzuki, JAEA. Introduction For the purpose of benchmarking the RANS code, FK- and experiments conducted at NSRR were analyzed. Emphasis was placed
More informationONR-RRR-088 Revision 0. Research Project ONR-RRR-088
Title of document RESEARCH REPORT Unique Document ID and Revision No: ONR-RRR-088 Revision 0 Project: Title: Research Project ONR-RRR-088 Review of the iron-based materials applicable for the fuel and
More informationDry storage systems and aging management
Dry storage systems and aging management H.Issard, AREVA TN, France IAEA TM 47934 LESSONS LEARNED IN SPENT FUEL MANAGEMENT Vienna, 8-10 July 2014 AREVA TN Summary Dry storage systems and AREVA Experience
More informationExperimental irradiations of materials and fuels in the BR2 reactor
Experimental irradiations of materials and fuels in the BR2 reactor Steven Van Dyck Co-authored by E. Koonen, M. Verwerft, M. Wéber IAEA technical meeting on Commercial products and services of research
More informationUtilization of Serpent for practical applications at Fortum
8th International Serpent UGM, Espoo, Finland Utilization of Serpent for practical applications at Fortum 1.6.2018 Jaakko Kuopanportti, Tuukka Lahtinen Fortum Power and Heat Oy Content Intro Comparison
More informationChallenges of structural materials for innovative nuclear systems in Europe
Challenges of structural materials for innovative nuclear systems in Europe Marta Serrano, Dolores Gomez-Briceño Structural Material Division CIEMAT Joint EC-IAEA Topical Meeting on Development of New
More informationThermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup
M.H. Altaf and Atom N.H. Badrun Indonesia / Atom Vol. 40 Indonesia No. 3 (2014) Vol. 40107 No. - 112 3 (2014) 107-112 Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering
More informationChapter 4 THE HIGH TEMPERATURE GAS COOLED REACTOR TEST MODULE CORE PHYSICS BENCHMARKS
Chapter 4 THE HIGH TEMPERATURE GAS COOLED REACTOR TEST MODULE CORE PHYSICS BENCHMARKS 4.1 HTR-10 GENERAL INFORMATION China has a substantial programme for the development of advanced reactors that have
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
DIMENSIONAL BEHAVIOUR TESTING OF ACCIDENT TOLERANT FUEL (ATF) IN THE HALDEN REACTOR R. Szőke, M. A. McGrath, P. Bennett Institute for Energy Technology OECD Halden Reactor Project ABSTRACT In order to
More informationNeutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code
Journal of Physics: Conference Series PAPER OPEN ACCESS Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code To cite this article: K Tiyapun and S
More informationStatus of Gen-IV and SMR Development Today
25 th SKC Symposium October 10 11, 2017 Fagerudd konferens, Enköping KTH ROYAL INSTITUTE OF TECHNOLOGY Status of Gen-IV and SMR Development Today Sara Bortot OUTLINE INTRODUCTION AND OVERVIEW GENERATION-IV
More informationNuclear Materials Research in EU: present status and future perspectives Concetta Fazio
Seminar on Generation IV Nuclear Energy Systems Risø, Denmark 29-31 October, 2012 Nuclear Materials Research in EU: present status and future perspectives Program Nuclear Saftey Research KIT Universität
More informationDEVELOPMENT OF EDUCATION AND TRAINING PROGRAMS USING ISIS RESEARCH REACTOR
A5 DEVELOPMENT OF EDUCATION AND TRAINING PROGRAMS USING ISIS RESEARCH REACTOR F. FOULON, B. LESCOP National Institute for Nuclear Science and Technology, CEA, Saclay Research Center, Gif-sur-Yvette, France
More informationPROPOSED SUB-CRITICALITY LEVEL FOR AN 80 MW TH LEAD-BISMUTH-COOLED ADS
PROPOSED SUB-CRITICALITY LEVEL FOR AN 80 MW TH LEAD-BISMUTH-COOLED ADS L. Mansani, R. Monti and P. Neuhold Ansaldo Nuclear Division, C.so Perrone, 25, 16161 Genova, Italy, Mansani@ansaldo.it Abstract The
More informationEnhanced Accident Tolerant Fuel at AREVA NP. Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017
Enhanced Accident Tolerant Fuel at Dr. Elmar Schweitzer, Dr. Jeremy Bischoff COP23, Bonn, 11/08/2017 Why Develop eatf Solutions? Zr alloy eatf solution p.2 eatf Program u Evolutionary Concept (Near-term
More informationThe Next Generation Nuclear Plant (NGNP)
The Next Generation Nuclear Plant (NGNP) Dr. David Petti Laboratory Fellow Director VHTR Technology Development Office High Temperature, Gas-Cooled Reactor Experience HTGR PROTOTYPE PLANTS DEMONSTRATION
More informationTHE KNOWN UNKNOWNS OF MOLTEN SALT REACTORS
THE KNOWN UNKNOWNS OF MOLTEN SALT REACTORS R. Ortega Pelayo 1, M. Edwards 2 1 Canadian Nuclear Laboratories,Chalk River, Ontario, Canada (286 Plant Road, Stn. 42, (613) 584 3311 ext. 44155, rosaelia.ortegapelayo@cnl.ca)
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
FRESH WATER QUENCHING OF ALLOYS OF NUCLEAR INTEREST INCLUDING FeCrAl FOR ACCIDENT TOLERANT FUEL CLADDING Michael Schuster 1, Cole Crawford 1, and Raul B. Rebak 1 1 GE Global Research: 1 Research Circle,
More informationTransmutation of Transuranic Elements and Long Lived Fission Products in Fusion Devices Y. Gohar
Transmutation of Transuranic Elements and Long Lived Fission Products in Fusion Devices Y. Gohar Fusion Power Program Technology Division Argonne National Laboratory 9700 S. Cass Avenue, Argonne, IL 60439,
More informationNEUTRONIC ASSESSMENT ON THE USE OF ADVANCED COATED PARTICLES IN A FLUIDIZED BED NUCLEAR REACTOR
NEUTRONIC ASSESSMENT ON THE USE OF ADVANCED COATED PARTICLES IN A FLUIDIZED BED NUCLEAR REACTOR Alexander Agung Department of Nuclear Engineering and Engineering Physics, Faculty of Engineering, Universitas
More informationFinal Results: PWR MOX/UO 2 Control Rod Eject Benchmark
Final Results: PWR MOX/UO 2 Control Rod Eject Benchmark T. Kozlowski T. J. Downar Purdue University January 25, 2006 This work has been sponsored by the U.S. Nuclear Regulatory Commission. The views expressed
More informationJAEA s Activities on Nitride Fuel Research for MA Transmutation
9 th IEM on Actinide and Fission Product Partitioning and Transmutation 26-28 September 2006 Hotel Novotel Atria, Nîmes, France JAEA s Activities on Nitride Fuel Research for MA Transmutation Yasuo ARAI,
More informationSupercritical Water Reactor Review Meeting. Materials Issues
Supercritical Water Reactor Review Meeting Materials Issues Bill Corwin, Louis Mansur, Randy Nanstad, Arthur Rowcliffe, Bob Swindeman, Peter Tortorelli, Dane Wilson, Ian Wright Oak Ridge National Laboratory
More informationCalculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol. 2, pp.301-305 (2011) TECHNICAL MATERIAL Calculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes Motomu SUZUKI *, Toru
More informationBN-1200 Reactor Power Unit Design Development
International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13) BN-1200 Reactor Power Unit Design Development B.A. Vasilyev a, S.F. Shepelev a, M.R.
More informationPhysical Properties. Can increase the strength by cold working but the recrystallization temperature is 400 to 500 C
Zirconium Cladding Why? Physical Properties Corrosion Resistance Radiation Effects ----------------------------------------------- In the early 1950Õs the Navy was looking for a material with low σ a high
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
WELD DEVELOPMENT OF FE-CR-AL THIN-WALL CLADDING FOR LWR ACCIDENT TOLERANT FUEL Jian Gan 1, Nathan Jerred 1,2, Emmanuel Perez 1, DC Haggard 2, Haiming Wen 1,3 1 Idaho National Laboratory: 1625 PO Box, Idaho
More informationDEVELOPMENT OF ADVANCED MIXED OXIDE FUELS FOR PLUTONIUM MANAGEMENT
LA-UR-97-2462 June 1997 DEVELOPMENT OF ADVANCED MIXED OXIDE FUELS FOR PLUTONIUM MANAGEMENT Stacey Eaton, Carl Beard, John Buksa, Darryl Butt, Kenneth Chidester, George Havrilla, and Kevin Ramsey DEVELOPMENT
More information