Report EUR 25326 EN

2012

Edited by A. Zeman (IAEA) P. Hähner (Institute for Energy and Transport, JRC/EC)

Book of abstracts 2nd joint IAEA-EC topical meeting

Development of new structural materials for advanced fission and fusion reactor systems

European Commission Joint Research Centre

Institute for Energy and Transport

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JRC70758

EUR 25326 EN

ISBN 978-92-79-24908-2 (pdf)

ISBN 978-92-79-24907-5 (print)

ISSN 1831-9424 (online)

ISSN 1018-5593 (print)

doi:10.2790/52879

Luxembourg: Publications Office of the European Union, 2012

© European Union, 2012

Reproduction is authorised provided the source is acknowledged.

DEVELOPMENT OF NEW STRUCTURAL MATERIALS FOR

ADVANCED FISSION AND FUSION REACTOR SYSTEMS

BOOK of ABSTRACTS

2ND JOINT IAEA-EC TOPICAL MEETING

in cooperation with the

16 – 20 April 2012, JRC Ispra (Italy)

Edited by A. Zeman (IAEA) and P. Hähner (JRC-Institute for Energy and Transport)

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MEETING VENUE

JRC Ispra Amphitheatre European Commission Joint Research Centre

Institute for Energy and Transport Renewable Energy Unit

Via E. Fermi, 2749 IT-21027 Ispra (VA), Italy

LOCATION:

http://ies.jrc.ec.europa.eu/the-institute/location-2.html

LOCAL CONTACT:

Ms Willy Muntjewerf European Commission Joint Research Centre

Institute for Energy and Transport F04 Unit

P.O.Box 2 NL-1755 ZG Petten

Tel: +31 653 776 233 willy.muntjewerf@ec.europa.eu

Ms Rita Sofia Paiola

European Commission Joint Research Centre

Institute for Energy and Transport Renewable Energy Unit

Via E. Fermi, 2749 IT-21027 Ispra (VA), Italy

Tel: +39 0332 78 9733 Fax:+39 0332 78 5869

rita.paiola@ec.europa.eu

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Foreword

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Further to a successful previous Topical Meeting on the "Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems" jointly organised by IAEA and JRC-Institute for Energy and Transport and held in Oct. 2009 at the premises of Fusion for Energy in Barcelona, the 2nd Joint IAEA-EC Topical Meeting on the same subject took place on 16 – 20 April 2012 at the JRC in Ispra, Italy. The Topical Meeting has again provided a well received platform for detailed presentations, technical discussions and exchange of results in the specific areas of relevance to materials performance assessment and qualification for advanced fission on the one hand, and thermo-nuclear fusion systems on the other hand. In fact, the Topical Meeting has achieved its objective to gather experts from both scientific communities, in order to develop synergies between the fields of research. Following keen demand for participation, the Topical Meeting was limited to 65 international delegates from 20 countries, with the strongest participation from the USA, Russia, Japan, Germany, France, and The Netherlands. The present report contains the collection of abstracts of the papers presented, while full size papers will be published as a Themed Issue of Journal of Nuclear Materials.

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Committees

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GENERAL CHAIR

S.ZINKLE ORNL (USA)

_________________________

PROGRAM COMMITTEE

E. DIEGELE F4E (ESP)

P. HAEHNER IET-JRC-EC (NL)

A. ZEMAN IAEA (AUT)

_________________________

SCIENTIFIC COMMITTEE

T. ALLEN WISC. U. (USA)

M. BRUMOVSKY NRI (CZ)

S. C.CHETAL IGCAR (IND)

S. DUDAREV CCFE (UK)

C. FAZIO KIT (GER)

N. GHONIEM UCLA (USA)

D. HOELZER ORNL (USA)

W. HOFFELNER PSI (SWI)

A. KIMURA KYOTO U. (JAP)

A. RYAZANOV RRC-KI (RUS)

T. SHIKAMA TOHOKU U. (JAP)

V. SLUGEN STU (SVK)

F. TAVASSOLI CEA (FRA)

R. VILA CIEMAT (ESP)

V. VOYEVODIN KIPT (UKR)

_________________________

ORGANISATION COMMITTEE

W. MUNTJEWERF IET-JRC-EC (NL)

R. PAIOLA IET-JRC-EC (ITA)

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Programme

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2nd Joint IAEA-EC Topical Meeting Monday 16 AprilLocation : Amphitheatre, JRC IspraIntroduction11h00 - 11h10 Welcome note by IET-JRC Director Page11h10 - 11h30 Zeman Openning remarks - IAEA activities O-1 p.1811h30 - 11h50 Haehner Openning remarks - JRC/EC activities O-211h50 - 12h30 Zinkle Pathways for Improving Nuclear Reactor Structural Materials Performance K-1 p.2212h30 - 14h00 LunchQualification I (Session chair: S.Zinkle)14h00 - 14h30 Fluss Retiring Risk and Qualification for Inertial Fusion Energy Structural Material K-2 p.2314h30 - 14h50 Tavassoli Updating Stainless Steel 316L(N) Properties for Application in Fission and Fusion C-1 p.4714h50 - 15h10 Jayakumar Reduced Activation Ferritic Martensitic Steel and Fabrication

Technologies for India's Testing Blanket Module (TBM) in ITER C-2 p.4815h10 - 15h30 Luzginova Fracture Properties of Neutron Irradiated ODS Eurofer97 Steel C-3 p.5015h30 - 16h00 Coffee breakQualification II (Session chair: J.Park)16h00 - 16h30 Chetal Development of Materials and Fabrication Technologies for Sodium

Cooled Fast Reactor Systems K-3 p.2516h30 - 16h50 Gavrilov Material Issues for Design and Licensing of MYRRHA ADS System C-4 p.5116h50 - 17h10 Riley Development of High Temperature Materials for Nuclear Applications C-5 p.52

17h10 17h30Chaouadi/ Gavrilov Effect of crack length-to-width ratio on crack resistance of high Cr ODS steels C-6 p.53

17h30 - 18h30 IAEA CRP Satellite meeting on ODS distribution (CRP members only) 2nd Joint IAEA-EC Topical Meeting Tuesday 17 AprilLocation : Amphitheatre, JRC IspraNew steels I (Session chair: J. de Carlan)9h00 - 9h30 Toualbi

K-4 p.279h30 - 9h50 Jang Crystallographic Relationship of Oxide Particles with Matrix in 12Cr

ODS Steel C-7 p.549h50 - 10h10 Kim, J. Microstructures and Strengthening Mechanisms of a Nanostructured

Ferritic Alloy at Wide Temperatures C-8 p.5510h10 - 10h30 Serrano Evaluation of Mechanical Properties of ADS Alloys in the GETMAT Project C-9 p.5610h30 - 11:00 Coffee breakNew steels II (Session chair: D.Hoelzer)11h00 - 11h30 Park Development of Nano-Sized Ceramic Dispersed Carbon Steel by

Conventional Casting Process K-5 p.2911h30 - 11h50 Gillemot Low Activation CrMoV Steel C-10 p.5711h50 - 12h10 Hoffmann Investigation on Different Oxides as Candidates for Nano-Sized ODS

Particles in Reduced-Activation Ferritic (RAF) Steels C-11 p.5812h10 - 12h30 Kytka/ Kopriva Irradiation embrittlement characterization of the EUROFER97 material C-12 p.5912h30 - 14h00 LunchTechnical issues for GenIV and DEMO I (Session chair: A.Zeman)

14h00 - 14h30 AntuschTwo Component Tungsten Powder Injection Molding - An Effective Mass Prod. Process K-6 p.30

14h30 - 14h50 Hoelzer Feasibility of Joining of 14YWT and F82H by Friction Stir Welding C-13 p.6114h50 - 15h10 Li Study on Electron Beam Welding and PWHT for CLAM Steel C-14 p.6215h10 - 15h30 Frayssines Microstructure and Mechanical Properties of Optimised CuCrZr Alloy

after HIP C-15 p.63

15h30 - 16h00 Coffee breakTechnical issues for GenIV and DEMO II (Session chair: S.Chetal)16h00 - 16h30 Gonzalez Development of ODS RAF Steels under EFDA K-7 p.3116h30 - 16h50 Commin Fail Safe and Cost Effective Fabrication of a First Wall by Diffusion

Welding C-16 p.6416h50 - 17h10 Ryazanov Investigations of radiation damage on behaviour of structural and

plasma facing materials for fusion reactors C-17 p.6517h10 - 17h30 Chikada Deuterium Permeation through Erbium Oxide Coating on RAFM Steels

by Dipcoating Technique C-18 p.6717h30 - 18h00 Diegele /

GonzalezStructural Materials for DEMO: Options, R&D Status, issues andprospects

K-8 p.32

Relationships between Mechanical Behavior and Microstructural Evolutions in Fe 9/14Cr-ODS during the Fabrication Route of SFR Cladding Tubes

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2nd Joint IAEA-EC Topical Meeting Wednesday 18 AprilLocation : Amphitheatre, JRC IspraNew steels III (Session chair: A.Ryazanov)9h00 - 9h30 De Carlan Understanding and First Modelling of the Tensile Behaviour of an ODS

Ferritic Alloy K-9 p.339h30 - 9h50 Sokolov Fracture Toughness of Nanostructured Ferritic Alloys for Nuclear Applications C-19 p.699h50 - 10h10 Yamashita Irradiation Behaviour Evaluation of Oxide Dispersion Strengthened

Ferritic Steel Cladding Tubes Irradiated in JOYO C-20 p.7010h10 - 10h30 Rogozkhin Tomographic atom probe study of un- and irradiated ODS Eurofer steel C-21 p.7110h30 - 11:00 Coffee breakCoolant compatibility and advanced testing I (Session chair: R.Novotny )11h00 - 11h30 Ignatiev Tellurium Corrosion of Nickel-Based Alloys in Molten Salt Fast Reactor Fuels K-10 p.3411h30 - 11h50 Novotny Miniaturised Material Testing Devises for Super Critical Water (SCW) Environment C-22 p.7211h50 - 12h10 Tsisar Effect of Liquid Metals (Pb, Li) on the Mechanical and Corrosion

Properties of Low Activation Materials as Applied for Fission and Fusion Reactor Concepts C-23 p.73

12h10 - 12h30 Slugen Evaluation of RPV steels by positron annihilation C-24 p.7512h30 - 14h00 Lunch14h00 - SOCIAL PROGRAMME

20h00 DINNER

2nd Joint IAEA-EC Topical Meeting Thursday 19 AprilLocation : Amphitheatre, JRC IspraRadiation damage I (Session chair: M.Fluss )9h00 - 9h30 Hsiung Cavity Formation in Multi-Ion-Beam Irradiated ODS Ferritic Steel K-11 p.369h30 - 9h50 Hernandez-MayoEvolution of the Microstructure of FeCr Alloys under Neutron and Ion Irradiation C-25 p.769h50 - 10h10 Al Mazouzi Prediction of the effects of Radiation for reactor pressure vessel and in-

vessel materials using multi-scale modelling C-26 p.7810h10 - 10h30 Kryukov Prediction of Irradiation Embrittlement of Vanadium Alloyed Low Nickel

Steel for Future Reactors C-27 p.7910h30 - 11:00 Coffee breakCoolant compatibility and advanced testing II (Session chair: P.Haehner )11h00 - 11h30 Yeliseyeva Corrosion Behaviour of ODS Steels in the Lead Melts Doped by

Oxygen K-12 p.3711h30 - 11h50 Moilanen Pneumatically powered Material Testing Devices for Fusion and GEN

IV Applications C-28 p.8011h50 - 12h10 Nishimura A New Test System of Axial Strain Controlled Fatigue Test with

Miniature Specimen C-29 p.8112h10 - 12h30 Blagoeva Stability of Ferritic Steel to Higher Doses; Survey of RPV Data and

Comparison with Candidae Materials for Future Systems C-30 p.8212h30 - 14h00 LunchCeramics, refractory metals I (Session chair: S.Gonzalez )14h00 - 14h30 Ferraris Joining and Integration Issues of Ceramics and CMC for Nuclear Applications K-13 p.3814h30 - 14h50 Capriotti /

ManaraExperimental Characterization of the Melting of Refractory Oxides: A Laser Heating Investigation of Nuclear Materials C-31 p.83

14h50 - 15h10 Gentzbittel Assessment of Vanadium Alloys for use in Fission Reactors C-32 p.8415h10 - 15h30 Sivak

C-33 p.8515h30 - 16h00 Coffee breakRadiation damage II (Session chair: T.Jayakumar )16h00 - 16h30 Bergner

K-14 p.3916h30 - 16h50 Terentyev/

MalerbaC-34

p.8616h50 - 17h10 Sekio C-35

p.8717h10 - 17h30 Voyevodin C-36 p.8817h30 - 18h00 Gilbert Integrated Materials Lifetime Assessment for DEMO Neutron

Transport, Inventory and Atomistic Calculations K-15 p.41

Use of self-ion irradiation to study void swelling and phase stability in advanced

The Effects of Irradiation Defect Distribution and the Steel Compositions on Void Denuded Zone Formations during Neutron Irradiation and Electron Irradiations

Dislocation Sinks Efficiency under Different Temperatures and Applied Loads inBCC Iron and Vanadium

Critical Consideration of the Composition of Cr-Rich Clusters in Neutron-Irradiated Fe12at%CrMicrochemical effects in irradiated Fe-Cr alloys as revealed by atomistic simulation

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2nd Joint IAEA-EC Topical Meeting Friday 20 AprilLocation : Amphitheatre, JRC IspraRadiation damage III (Session chair: L.Malerba )9h00 - 9h30 Garner Void Swelling and Irradiation Creep of Ferritic-Martensitic Alloys at Very

High DPA Levels Produced by Neutron and Self-Ion Irradiation K-16 p.429h30 - 9h50 He Synergistic effect of He and displacement cascade in FeCr alloys

studied at atomistic scale C-37 p.899h50 - 10h10 Ogorodnikova Effect on Radiation-Induced Damage on Deuterium Retention in

Tungsten and Tungsten Coatings C-38 p.9010h10 - 10h30 Shikama Study on Heavy Irradiation Damage in Maerials C-39 p.9110h30 - 11:00 Coffee breakRadiation damage IV (Session chair: F.Garner )11h00 - 11h30 Malerba Effect of Dislocation Loop Decoration on Radiation-Induced

Hardening in Different Iron Alloys K-17 p.4311h30 - 11h50 Skuratov Radiation Stability of Oxide Nanoslusters in ODS Alloys Against Swift

Heavy Ion Impact C-40 p.9211h50 12h10 van Vuuren Radiation tolerance of nanostructured ZrN coatings against swift

heavy ion irradiation C-41 p.9312h10 - 12h30 Lindau Investigations on the Joining of 9-20Cr ODS and Non-ODS steels

applying Diffusion, Electron Beam and Friction Stir Welding C-42 p.9412h30 - 12h45 Conclusions12h45 - 14h00 Lunch + END OF MEETING

14:00 Transport to MILANO MALPENSA Airport

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Opening Papers

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O-1

IAEA COORDINATED RESEARCH ACTIVITIES ON STRUCTURAL MATERIALS FOR INNOVATIVE FUSION AND FISSION REACTORS

A.Zeman1*, R.L.Beatty2, R.Kaiser1, R.Kamendie1, M.Venkatesh1

1Department of Nuclear Sciences and Applications 2Department of Nuclear Energy

International Atomic Energy Agency, Vienna, Austria *Email: azeman@iaea.org

The performance and integrity of structural and functional materials are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy applications, specifically fission and fusion reactors, fuel recycling and waste transmutation systems or fusion-fission hybrid reactors. In view of international cooperation, there are two main on-going initiatives: (1) Generation IV International Forum (GIF) and (2) International Project for Innovative Reactors and Fuel Cycles (INPRO), both have been active for almost a decade. The development of new and innovative nuclear reactors and fuel cycles will be a key pillar in further strengthening the safe use of nuclear energy in the 21st century. Under INPRO, a new and dedicated project will study nuclear safety issues of innovative nuclear reactors and fuel cycles using the INPRO methodology including the consideration of combined multiple simultaneous natural and technical disasters or human error. The project will evaluate advanced nuclear energy systems and technologies that promise to meet the requirement that a major release of radioactivity will not happen in case of an accident, so that there is no need for evacuation measures outside the plant site. Meeting this requirement is crucial to obtain public acceptance and for the sustainability of nuclear energy. In line with the statutory objective expressed in Article II (The Agency shall seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world. It shall ensure, insofar as it is able, that assistance provided by it or at its request or under its supervision or control is not used in such a way as to further any military purpose), the IAEA is assisting Member States’ activities in the area of advanced reactor research and technology development by providing an umbrella for information exchange and collaborative R&D to pool/share resources and expertise. Nevertheless successful deployment of the new nuclear reactor systems will definitely require the development, qualification and deployment of new structural materials with improved mechanical properties combined with radiation and corrosion resistance. Although, there have been promising developments in new classes of advanced materials such as fibre-reinforced ceramic composite materials, oxide dispersion strengthened alloys with nano-structured features or advanced reduced activation ferritic-martensitic steels, sufficient radiation resistance, good corrosion, strength and toughness properties at high-temperatures are still quite challenging features.

It is clear that a number of complex experimental tests are urgently required in order to provide reliable and validated data and information which will lead to the qualification and licensing of materials. This includes the comprehensive in-pile irradiation programmes. Many of these activities require long-term intensive and internationally collaborative and coordinated programmes of research and development in which the IAEA can play an important role in the facilitation and coordination process. In principle these programmes aim to enhance the

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capability of interested Member States to build up advanced or innovative nuclear technologies by promotion of information exchange and collaborative R&D to resolve issues associated with specific problems. Generally, the role of the IAEA is to provide an appropriate forum for international dialogue and coordination of individual national initiatives in order to accelerate and boost research efforts. This broader initiative will accelerate technological improvements of the materials and design methodologies. The main beneficiaries of the internationally coordinated projects are the end-users from the participating Member States who will have a chance to implement the technology results that are achieved during the projects. Specifically, the activities described in this paper are carried out jointly by the Department of Nuclear Sciences and Applications (NA) and Department of Nuclear Energy (NE).

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K-1

PATHWAYS FOR IMPROVING NUCLEAR REACTOR STRUCTURAL MATERIALS PERFORMANCE

S.J. Zinkle

Oak Ridge National Laboratory,

P.O. Box 2008, Oak Ridge, TN 37831-6249 USA Email: zinklesj@ornl.gov

Future fission and proposed fusion energy systems will be increasingly dependent on advanced structural materials to reliably deliver high performance with favorable safety attributes and acceptable economic cost. In many cases, the proposed operating temperatures are significantly higher than the experience base for light water reactors. This motivates development of structural materials with improved high temperature strength for prolonged operating periods and engineered corrosion resistance for the candidate coolants and other materials in the system. The high radiation fluxes in future nuclear energy systems will require the structural materials to have superior radiation resistance compared to currently available materials. This presentation will review some of the current and emerging strategies being employed to develop structural materials with simultaneous high radiation resistance, high strength, good toughness and corrosion resistance, and moderate fabrication cost. Materials systems to be covered include steels, refractory alloys, bulk amorphous metals, and ceramic composites. Engineering approaches to design improved performance include introduction of nanostructures in the form of high densities of second phases (precipitates or inert particles) or nanolayered interfaces, development of tailored composite materials systems, creation of high strength and radiation-stable amorphous alloys, and design of optimized engineered grain boundaries. In the future, utilization of advanced manufacturing processes to produce near-net shape parts with precise microstructural control will be of increasing importance to control fabrication costs and to create high-performance fabrication architectures that could not be achieved using conventional fabrication methods.

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Keynote Papers

(in order of the programme schedule)

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K-2

RETIRING RISK AND QUALIFICATION FOR INERTIAL FUSION ENERGY STRUCTURAL MATERIALS

M. J. Fluss1*, L.Hsiung1, J.Marian1, P.Hosemann2, J.Latkowski1, and

M.Dunne1

1Lawrence Livermore National Laboratory, Livermore, CA 2University of California, Berkeley, CA Department of Nuclear Engineering

*Email: fluss1@llnl.gov

The need to qualify structural materials to withstand long term exposure to a fusion environment – and do so without access to a fully representative offline testing facility – has been a major limiting factor for all types of fusion. Fusion deployment has had to await the development of new materials able to survive for the operational lifetime and total radiation dose of the power plant (as much as 200 times that of a fission reactor). While research on candidate materials has been underway for some time, their qualification requires new radiation facilities that may not exist for a decade or more. This has clearly limited the motivation and ability to build commercial fusion power plants in a timely manner. A design strategy for Laser Inertial Fusion Energy (LIFE) has been adopted that circumvents the need for materials to tolerate extraordinary high radiation dose in excess of 150 dpa but rather relies on a replaceable first wall and blanket that will accumulate only a fraction of the life-time dose. The strategy calls for the construction of a first test facility (Phase I engine) which will be designed to see a dose of ~10 dpa , and which will serve as a platform for accelerated materials testing in a fusion environment. By simplifying the chamber design, and employing solutions that allow the use of a moderately sized chamber, it is possible to replace the entirety of the first wall and blanket at regular intervals during the operational life of the engine. By reducing the required material lifetime from 40-60 years (as currently suggested for a sustained plasma engine) to <5 years, a viable route to material benchmarking, down selection, and eventual qualification is realizable. The path for materials selection must be such that risks are retired early in the design and development process, thus bounding the performance limits and reducing consequential uncertainties while minimizing technological and financial risk. A commercial scale Phase II engine that will employ advanced material for the first wall and blanket will follow the Phase I engine. The Phase I engine will be the platform for executing the material experiments and tests required for regulatory certification of the Phase II engine. What this means is that the early materials research before the construction of the Phase I engine is a critical precursor for the overall strategy which will lead to fusion environment testing of materials in the Phase I engine. In this presentation we will discuss a strategy of experiments and modeling to achieve an early down selection of materials for a Phase I IFE engine using ion-beams and eventually spallation neutrons. Some interesting experimental and modeling results from our ion-beam studies using dual and triple beam irradiations of Fe, Fe(Cr) and ODS materials will be highlighted. The advantages and the concerns of materials selection and qualification without an actual fusion source of neutrons will be discussed and some specific examples of each will be illuminated. ___________________

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Ion beam irradiations performed at CEA-Saclay, France, multi-ion-beam JANNuS Facility. This work performed in part under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344. Some of this work was coordinated with Accelerator Simulation and Theoretical Modeling of Radiation Effects, IAEA Coordinated Research Program – SMoRE (Vienna, Austria)

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K-3

DEVELOPMENT OF MATERIALS AND FABRICATION TECHNOLOGIES FOR SODIUM COOLED FAST REACTOR SYSTEMS

S.C. Chetal Indira Gandhi Centre for Atomic Research, Kalpakkam 601 102, India

Email: chetal@igcar.gov.in

A comprehensive research and development programme on materials development and fabrication technologies is being pursued at Indira Gandhi Centre for Atomic Research, Kalpakkam, India for harnessing energy from fast-neutron fission and fusion reactors. Developments of material and fabrication technology play crucial role in the economic feasibility of fast nuclear power plant. In order to meet these objectives one of the methods is to extend the fuel burnup and decreasing doubling time. The burnup is largely limited by the void swelling and creep resistances of the fuel cladding and wrapping materials. India’s Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are alloy D9 as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generators. In order to improve the burnup, titanium, phosphorous and silicon contents in alloy D9 have been optimized for better swelling and creep resistances to develop modified version of alloy D9 as IFAC-1. Creep resistance of inherently void swelling resistance 9Cr-ferritic steel has been improved with the dispersion nano-size yttria to develop oxide dispersion strengthened (ODS) clad tube with long-term creep strength equivalent to D9 for increasing the fuel burnup. Development of modified 9Cr-1Mo steel clad tubes and 9Cr-1Mo steel wrapper for future metallic fuel reactor for reducing the doubling time are in progress. Extensive studies on resistance of this new generation core materials to void swelling is also under progress along with material development. Improved versions of 316LN stainless steel with nitrogen increased up to 0.14 wt.% and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron have also been developed indigenously. Special welding consumable with stringent requirement of composition and resistance to embrittlement after prolonged exposure to high temperature, has been developed indigenously to weld 316LN stainless steels. Development of fully austenitic IFAC-1, with a composition highly susceptible to hot cracking, required development of special welding procedure to join the clad tube made of this steel with end plugs of 316LN steel. Similar development has been carried out for welding of P91 clad tube with end plug to use with metallic fuel. A procedure is also being developed to join the ferritic ODS clad tubes with end plugs made of modified 9Cr-1Mo steel. The challenges to produce 25 m long seamless tubes and tube sheet forgings of thickness of ~ 300 mm for steam generator meeting the specification requirements, and welding and inspection of the tube to tube sheet joints have been realized. India’s participation in ITER project has given us an opportunity to develop our own version of Reduced Activation Ferritic Martensitic (RAFM) steel for Test Blanket Module (TBM) to be tested in ITER. The 9Cr-RAFM steel with 1.4 wt.% W and 0.06 wt.% of Ta has been developed for better combination of creep and fatigue strength, ductility and toughness. Extensive characterisation of this steel including effect of irradiation is under progress. Welding consumables for joining this steel have also been successfully developed. Technology for joining of this steel using Hot

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Isostatic Pressing, Electron Beam Welding, Laser and Laser Hybrid welding and Gas Tungsten Arc Welding are being developed. This paper would provide an overview of these developments and include salient results from the extensive testing and characterisation that have been carried out on these materials.

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K-4

RELATIONSHIPS BETWEEN MECHANICAL BEHAVIOR AND MICROSTRUCTURAL EVOLUTIONS IN FE 9/14CR-ODS DURING THE

FABRICATION ROUTE OF SFR CLADDING TUBES L. Toualbi1*, P. Olier2, C. Cayron3, J. Ribis1, M.-H. Mathon4, E. Rouesne1, D.

Bossu2, D. Nunes2, J.-L. Béchade1, R. Logé5, Y. de Carlan1

1DMN/SRMA/LA2M, CEA Saclay 2DMN/SRMA/LTMEx, CEA Saclay

3DRT/LITEN, CEA Grenoble 4Laboratoire Léon Brillouin, CEA/Saclay

5Centre de mise en forme des matériaux, Mines-Paristech, UMR CNRS 7635 *Email: louise.toualbi@cea.fr

Oxide Dispersion Strengthened (ODS) ferritic/martensitic steels present superior radiation resistance compared with austenitic steels and high creep strength due to a reinforcement by the homogeneous dispersion of hard nano-sized particles (such as Y2O3 or YTiO). They are considered as reference materials for high burn up cladding tubes for future Sodium-Cooled-Fast Reactors. Two new ODS alloys, one martensitic and one ferritic, are developed at CEA Saclay to achieve the goals defined for GEN IV reactors. The aim of this paper is to present the relationships between mechanical behavior and microstructural evolutions in the Fe 9-14Cr-ODS during the fabrication route of tubes. Powders atomized by Aubert &Duval were mechanically alloyed under hydrogen by Plansee. They were consolidated by hot extrusion to obtain raw tubes and manufactured into tube cladding using cold-rolling process. ODS steels are usually characterized by a low ductility and a high hardness at room temperature. The cold-working passes have to be punctuated by intermediate heat treatments in order to soften the raw tube and avoid any damage in the course of manufacturing. The ferritic alloy is a Fe-14Cr-1W-0.3Ti-0.3Y2O3 ODS which does not present any phase transformation. Releasing the internal stresses induced by the cold-rolling process was made by mean of recovery annealing. Heat treatments up to 1250°C were necessary to decrease the hardness of the material and to cold work the tubes. No recrystallization was clearly seen on this 14Cr-ODS ferritic alloy. Microstructural observations seems to indicate that the softening induced by heat treatments is more due to a recovery and/or a grain growth during the high temperature treatments than a real recrystallization. The tensile properties obtained on final tubes are in the meantime less favorable than the one of the 9Cr-ODS martensitic steel. The Fe-9Cr-1W-0.2Ti-0.3Y2O3 ODS martensitic steel presents a large austenitic domain at high temperature which allows complete phase transformation from ferrite to austenite. CCT diagram shows high critical cooling rates. Mechanical tests indicate that 9Cr-ODS cold manufacturing is much easier/safer under the softened ferritic phase. After each intermediate heat treatment, α-γ phase transformation led to softened ferritic state characterized by an isotropic microstructure which is very favorable to further cold rolling. On the other hand, for the last heat treatment a higher cooling rate was needed to obtain a martensitic microstructure. Final cladding tubes present an isotropic microstructure with an Ultimate Tensile Strength close to 550 MPa at 650°C and significant ductility. Electron Backscatter Diffraction analysis, Transmission

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Electron Microscopy and Small Angle Neutron Scattering measurements are used to characterize the microstructures.

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K-5

DEVELOPMENT OF NANO-SIZED CERAMIC DISPERSED CARBON STEEL BY CONVENTIONAL CASTING PROCESS

J.Park*, S. Hong, E-K. Park, M-K.Lee and Ch-K. Rhee

Nuclear Materials Development Division, Korea Atomic Energy Research Institute, Daejeon, Korea

*Email: jinjupark@kaeri.re.kr

In the present work, carbon steel dispersed with nano-sized TiC ceramic particles was fabricated by conventional casting process. The liquid metal casting process is more economical than other available routes for metal matrix nano-composites (MMNCs) production. However, it is so difficult to disperse nano-sized ceramic particles uniformly into molten metal due to a poor wettability and a specific gravity difference between the ceramic particles and metal matrix. In order to solve these problems, the nano-sized (~50nm) TiC powders were first mechanically activated in two metal ones by using a very high speed planetary ball mill machine and then they filled into the carbon steel capsule. The capsule filled with mechanical activated powders was finally introduced to the molten metal in a vacuum. The microstructural changes were examined by OM and SEM. The distribution of nano-sized TiC particles in the carbon steel matrix was studied by FE-TEM. According to OM and SEM images, it was found that the grain size refinement of the matrix is achieved when nano-sized TiC particles were added. FE-TEM images revealed that spherical TiC particles with several tens of nanometers (below 50nm) were distributed uniformly in the carbon steel matrix. The present work therefore proposes that conventional casting process can be an effective way to fabricate ODS alloys.

30

K-6

TWO COMPONENT TUNGSTEN POWDER INJECTION MOLDING – AN EFFECTIVE MASS PRODUCTION PROCESS

S.Antusch*, V.Piotter, M.Müller, T.Weingärtner

Institute for Applied Materials; Materials Processing Technology,

Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

*Email: steffen.antusch@kit.edu Tungsten and tungsten-alloys are presently considered to be the most promising materials for Plasma Facing Components for future fusion power plants. The Karlsruhe Institute of Technology (KIT) divertor design concept for the future DEMO power plant is based on modular He-cooled finger units. Each 1-Finger module consists of many single parts e.g. the tungsten tile and the tungsten alloy thimble, for the whole divertor system more than 250,000 single parts are needed. The advantages of tungsten and tungsten alloys are the high melting point, low activation, low erosion rate and high thermal conductivity. But the brittleness and hardness of these materials make the fabrication by mechanical machining such as turning and milling very difficult, time and cost intensive. Therefore, the development of suitable mass production methods for divertor parts was needed. A time and cost effective near-net-shape forming process with the advantage of shape complexity, material utilization and high final density is Powder Injection Molding (PIM). This process was adapted and developed at KIT for tungsten and promising results have already been achieved. Two component tungsten powder injection molding as further development allows the joining of two different materials e.g. tungsten with a doped tungsten alloy, without brazing. This contribution describes the complete technological process of two component powder injection molding for tungsten materials and its application on producing real DEMO divertor parts. Characterization results of the finished parts e.g. microstructure, hardness, density and joining zone quality are discussed.

31

K-7

DEVELOPMENT OF ODS RAF STEELS UNDER EFDA

S.M. Gonzalez de Vicente

EFDA Close Support Unit Boltzmannstrasse 2, D-85748 Garching, Germany

Email: Sehila.Gonzalez@efda.org In Europe present activities within the European Fusion Development Agreement (EFDA) on ODS RAF steels aim at developing materials with good tensile and creep strength and sufficient ductility, especially in terms of fracture toughness and ductile-to-brittle transition temperature (DBTT), with a focus on obtaining an isotropic microstructure and therefore isotropic properties. The EFDA work programme on ODS RAF steels has been recently re-structured, being organized along the four following programmatic lines:

− Production and characterization of laboratory scale batches of nano-structured ODSFS (identification of the optimal chemical composition, fabrication route and thermal mechanical treatments combination).

− Production and characterization of industrial batches of nano-structured ODSFS (production at industrial or semi – industrial scale of optimised generation of ODSFS with high tensile and creep strength at higher temperatures, and sufficient ductility and fracture toughness for DEMO relevant conditions).

− Irradiation and post- irradiation characterization of produced nano-structured ODSFS (investigate the effect of dose and temperature on hardening and the stability of oxide particles in the grain boundaries).

− State of art of nano – structured ODSFS: Bibliographical review (evaluation of the results obtained by USA and Japan in developing this kind of materials).

32

K-8

STRUCTURAL MATERIALS FOR DEMO: OPTIONS, R&D STATUS, ISSUES AND PROSPECTS

E. Diegele1*, S.M. Gonzalez de Vicent2, M. Rieth3

1F4E, Josep Pla 2, B3, Torres Diagonal Litoral, Barcelona, Spain

2EFDA, Close Support Unit, Boltzmannstrasse 2, D-85748 Garching, Germany 3Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344

Eggenstein-Leopoldshafen, Germany *Email: eberhard.diegele@f4e.europa.eu

In Fusion development the next step ahead is ITER, the facility to demonstrate the scientific and technological feasibility of fusion power. In parallel to ITER construction and operation open technology issues need to be addressed and solved before building a demonstration reactor (DEMO). This includes in particular structural materials R&D for in-vessel components. The loading is determined by high heat flux and damage through high energy 14 MeV neutrons. At the First Wall this typically corresponds to 30-80 dpa for DEMO, approximately two orders of magnitude higher than in ITER. Materials development for breeder blankets and divertors, has a history of more than two decades. The service, loading conditions and the required properties, in combination with safety standards, create a unique set of specifications. In particular, social-economic demands of low level waste and low activation reduce the choice significantly to four classes of structural materials: (i) the reduced activation ferritic/martensitic (RAFM) steels, including nano-dispersion strengthened variants, (ii) the vanadium alloys, (iii) the tungsten alloys and (iv) SiC fibre reinforced ceramic composites. The main objective is to have DEMO materials and key fabrication technologies fully developed and qualified (for full DEMO life) within two decades. Nevertheless, a major part of the &&D task has to be completed much earlier. An additional issue for materials R&D is that the definition of a DEMO is a moving target, differs from country to country and is adjusted to respective fusion roadmaps. The paper discusses communalities and differences of fusion road map as well as what is partly a consequence, different requests and boundary conditions for the materials R&D. In particular, different material classes of existing materials and materials still under development will be analyzes with respect to R&D status, major issues to be solved and prospects for their use in the next generation facilities.

33

K-9

UNDERSTANDING AND FIRST MODELLING OF THE TENSILE BEHAVIOUR OF AN ODS FERRITIC ALLOY

M. Ratti1, Y. de Carlan1*, J. Garnier1, J. Malaplate1, F. Dalle1, Y. Bréchet2

1Nuclear Materials Department, CEA Saclay, 91191 Gif-sur-Yvette, France

2INPG - SIMAP BP 75 38402 St Martin d'Heres cedex, France *Email: yann.decarlan@cea.fr

ODS ferritic alloys (Oxide Dispersion Strengthened) are considered for the cladding of future Sodium Fast Reactors. To obtain a better understanding of the mechanical behavior of this type of alloys, a Fe-18CrTiY2O3 ferritic ODS alloy was produced. The mechanical alloying was performed by Plansee under hydrogen from powder produced by Aubert &Duval. The consolidation was carried out by hot extrusion at 1100°C trough a rectangular die. The material was tensile tested at 20°C and 650°C with load applied along three directions: longitudinal, transverse and 45° with respect to the extrusion direction. The tensile behavior was modelled taking into account all the microstructural parameters of the material (crystallographic texture, grain size and morphology, distribution of nano-oxides). The isotropic, kinematic and viscous stresses were quantified at 20°C and 650°C and the calculations were found to match experimental observations. At low temperature, hardening due to the presence of nano-oxides seems to be prevalent. In contrast, at high temperatures it is rather the very low grain sizes which seems to explain the high elastic limits observed. The differences in behaviour between sampling directions are also analyzed, and explained by the crystallographic and morphological texture of the material. _______________ Key words: ODS materials, mechanical behaviour, modelling, tensile test.

34

K-10 TELLURIUM CORROSION OF NICKEL-BASED ALLOYS IN MOLTEN

SALT FAST REACTOR FUELS

V.Ignatiev*, A. Surenkov , I.Gnidoy and O.Feynberg

NRC Kurchatov Institute, Moscow, Russian Federation *Email: ignatiev@vver.kiae.ru

In Russia, R&D on Molten Salt Reactor (MSR) are concentrated now on fast/intermediate spectrum concepts which were recognized as long term alternative to solid fuelled fast reactors due to their attractive features: strong negative feedback coefficients, easy in-service inspection, and simplified fuel cycle. Designs are available for Th-U breeding and for long lived actinides burning. Selection of the salt composition strongly depends on the specific design application. For all MSR concepts, materials selection in primary circuit is a very important issue. Last years, Kurchatov Institute with partners contribute to development of key technical solutions for promising Molten Salt Fast Reactor (MSFR) concepts without and with Th-U support, including combined materials compatibility & salt chemistry control in selected fuel salt environments at parameters simulating design operation. This paper summarizes results of an experimental studies conducted to understand the mechanism and to develop a means of controlling tellurium intergranular cracking (IGC) in the Ni – base alloys. The addition of a chromium telluride to fuel salt can be used to provide small partial pressures of tellurium simulating a reactor environment where tellurium appears as a fission product. Tellurium corrosion of Ni-based alloys in stressed and unloaded conditions studies was tested in LiF-NaF-BeF2 and LiF-BeF2-ThF4-UF4 fuel salts at temperatures up to 740oC with measurement of the salt redox potential. Following Hastelloy N-type modified alloys: HN80M-VI with 1.5% Nb; HN80MTY with 1% Al and MONICR with about 2% Fe were used for the study in the 15LiF-58NaF-27BeF2 corrosion facilities. Materials investigated in LiF-BeF2-ThF4-UF4 fuel salt included, in addition to mentioned above, high temperature HN80MTB (77Ni-7Cr-10Mo-6W) and EM-721 (65Ni-28W-7Cr) alloys. The IGC produced in Hastelloy N (or HN80M) -type alloys when exposed to this chromium telluride – molten 15LiF-58NaF-27BeF2 salt mixture can be reduced by adding niobium or aluminium to the Ni base alloy or by controlling the oxidation potential of the salt in the reducing range. It was shown that both Re and Y additions only slightly increase the alloy’s resistance to tellurium cracking. The alloy doped with Nb alone significantly increases IGC resistance. The alloy containing both Тi and Nb did not provide required resistance to tellurium corrosion. Addition of Mn gives a significant increase in alloy resistance to tellurium IGC. Five tests of alloys specimens with exposure time 250 hrs each in the 70LiF-7BeF2-21ThF4-2(UF4+UF3) fuel salt were done in the range of the [U(IV)]/[U(III) ] ratio varied from 1 to 100. After materials exposure in the fuel salt with the [U(IV)]/[U(III)] ratio from 1 to 50 there was revealed no traces of tellurium IGC on specimens’ surface. The strength characteristics of alloys and their structure were changed insignificantly after tests. These changes were apparently stipulated by alloy structure, temperature/time factor and mechanical loads in a greater extent, than impact of a fuel salt containing tellurium additives. All alloys investigated have a good ductility at high strength characteristics. As to ЕМ-721

35

alloy, the visible features of brittle disruption are observed both before and after exposure in the fuel salt containing chromium telluride. HN80MTY alloy is the most resistant to tellurium IGC of Ni-base alloys under study. According our evaluation its corrosion and mechanical properties can meet burner and breeder MSR requirements at temperatures up to 740°C.

36

K-11

CAVITY FORMATION IN MULTI‐ION‐BEAM IRRADIATED ODS

FERRITIC STEEL

L. Hsiung1*, M. Fluss1, S. Tumey1, Y. Serruys2, F. Willaime2, A. Kimura3

1Lawrence Livermore National Laboratory, Livermore, CA94551, U.S.A.

2Service de Recherches de Métallurgie Physique (CEA), Gif‐sur‐Yvette 91191,

France

3Kyoto University, Gokasho, Uji, Kyoto 611‐0011, Japan

Email: hsiung1@llnl.gov

One of the critical challenges in designing fusion power reactors is to develop high performance structural materials for first wall and blanket components,

which will be exposed to an intense high‐energy (14 MeV) neutron flux and

helium (He) and hydrogen (H) transmutation gases. The intense neutron flux will generate large numbers of point defects which give rise to cavity formation. The process of cavity formation can be further promoted in the presence of helium and hydrogen gases in stimulating the formation of large helium bubbles and voids which cause void swelling and lead to dimensional changes. The formation of bubbles and voids at grain boundaries can potentially cause premature failure and reduce the lifetime of the first wall/blanket components of a fusion reactor. In this presentation, critical results generated from HRTEM

studies of (Fe + He) dual‐beam and (Fe + He + H) triple‐beam irradiated

Fe‐14Cr alloy and Fe‐16Cr ODS steel are reported. The results reveal that ODS

37

steel containing high‐density oxide nanoparticles can effectively trap

helium‐filled cavities at the particle‐matrix interfaces and significantly suppress

the void swelling and grainboundary cavitation under irradiation. This work was performed under the auspices of the U.S. Department of Energy by Lawrence

Livermore National Laboratory under Contract DE‐AC52‐07NA27344.

38

K-12

CORROSION BEHAVIOR OF ODS STEELS IN THE LEAD MELTS DOPED BY OXYGEN

O.Yeliseyeva1*, V.Tsisar1, Z.Zhou2

1Physical-Mechanical Institute of NASU, 5, Naukova St., L`viv 79601, Ukraine

2School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083, China

Email: olgayeliseyeva@ipm.lviv.ua The oxide dispersion strengthened (ODS) steels are considered as a candidate materials dedicated to increase the upper temperature limit of operation of fast reactors and accelerator driven systems cooled by heavy metal melts (Pb, Pb-Bi). Compatibility of named steels with the liquid metals is one of the main issues. This work is aimed at corrosion behavior of ferritic and ferritic-martensitic ODS steels in the liquid lead depending on temperature and the oxygen concentration in Pb. The corrosion tests were carried out at 550 and 650ºC for 1000 h in the static lead melt with various oxygen concentration: pure lead Pb-1 (≤5•10-7 wt%O); oxygen controlled Pb-2 (10-5-10-6 wt%O) and oxygen-saturated Pb-3 (~10-3 wt %O). It was determined that the corrosion behavior of both steels changed from dissolution in the lead with low oxygen (Pb-1) to formation on the steel’s surface of protective oxide layers (Pb-2) and finally to catastrophic oxidation in the lead saturated by oxygen (Pb-3). In general, that coincides with corrosion behavior of conventional chromium steels. As a contrast to conventional steels, the fine-grained structure of ODS steel caused intergranular porosity and selective dissolution of chromium in the pure lead (Pb-1) especially at 550ºC. At higher temperature (650ºC) dissolution of both Fe and Cr occurred. On the other hand in the melt with higher oxygen content (Pb-2), the fine-grained structure of ODS promoted diffusion of chromium along grain boundaries and therefore formation of protective oxide layer on the surface of steel. At low temperature (550ºC) the protective layer possessed double structure: outer magnetite and inner spinel layers. When the temperature of exposure reached 650º C the monolayer with high chromium content was formed on the surface of both steels. In the oxygen-saturated melt (Pb-3) the both steels underwent severe oxidation with formation of non-protective porous multiphase scale with Pb inclusions. Based on the experimental data the mechanisms of component interaction in the “steel - lead melt” system depending on temperature and oxygen content are proposed.

39

K-13

GLASS-CERAMICS AS JOINING MATERIALS FOR NUCLEAR APPLICATIONS

M. Ferraris

Politecnico di Torino, Corso Duca degli Abruzzi 24, Torino 10129, Italy

Email: monica.ferraris@polito.it Joining of SiC/SiC composites for nuclear applications can be of interest for both future thermo-nuclear fusion reactors and new generation fission reactors components. In both cases, the main issues are the extreme thermo-mechanical loads on the joined components, the not completely known service conditions and requirements, their resistance to high temperatures, to neutron irradiation and to harsh chemical environment. Three silica and non silica based glass-ceramic have been used to join SiC and SiC/SiC by a pressure-less joining technique. The mechanical characterization of the joints will be discussed in terms of bending and torsion tests, in particular for the latter, on joined miniaturized hour-glass shaped specimens, designed to fit irradiation capsules. Bending strength has been measured on glass-ceramic joined SiC/SiC before and after neutron irradiation at the High Flux Reactor in Petten: preliminary results showed almost the same bending strength before and after irradiation at 600 °C. Finally, the behaviour of glass-ceramics as joining materials for SiC/SiC will be shown before and after fast neutron irradiation: results showed the same glass ceramic structure and phases before and after irradiation at 600 °C. References [1] M. Ferraris, M. Salvo, V. Casalegno, S. Han, Y. Katoh, H.C. Jung, T. Hinoki,

A. Kohyama, “Joining of SiC-based materials for nuclear energy applications” , J. Nucl. Mat., in press, doi:10.1016/j.jnucmat.2010.12.160 (2011).

[2] L.L. Snead, T. Nozawa , M. Ferraris , Y. Katoh , R. Shinavski , M. Sawan, “Silicon carbide composites as fusion power reactor structural materials “ J. Nucl. Mater. (2011), in press doi:10.1016/j.jnucmat.2011.03.005.

[3] M. Ferraris, M. Salvo, V. Casalegno, S. Rizzo, A. Ventrella, “Joining and Integration Issues of Ceramic Matrix Composites for Nuclear Applications” in Processing and Properties of Advanced Ceramics and Composites II , Ceramic Transactions (2) Edited by Narottam P. Bansal, Jitendra P. Singh, Jacques Lamon Sung, R. Choi, Morsi M. Mahmoud (2010).

[4] M. Ferraris, M. Salvo, C. Isola, M. Appendino Montorsi, A. Kohyama, Glass ceramic joining and Coating of SiC/SiC for Fusion Applications , J.Nucl.Mat.258-263 1546-1550 (1998).

[5] Y. Katoh, M. Kotani, A. Kohyama, M. Montorsi, M. Salvo, M. Ferraris, “Microstructure and Mechanical Properties of Low activation glass-ceramic joining and Coating for SiC/SiC Composites” J.Nucl. Mater., 283-287 , 1262-1266 (2000)

40

K-14

CRITICAL CONSIDERATION OF THE COMPOSITION OF CR-RICH CLUSTERS IN NEUTRON-IRRADIATED FE-12AT%CR

F. Bergner1, V. Kuksenko2, L. Malerba3, C. Pareige2, P. Pareige2, A.

Ulbricht1, A.Wagner1

1HZDR, P.O.Box 510119, Dresden 01314, Germany 2CNRS, France;

3SCK•CEN, Boeretang 200, Mol 2400, Belgium *Email: f.bergner@hzdr.de

The composition of solute-enriched clusters and precipitates formed in Fe-Cr alloys as the result of neutron irradiation is an unsolved issue. It is an important issue for several reasons, namely

− to reach a complete and consistent description of the nanoscale features derived from the application of necessarily several complementary techniques,

− to correctly design and calibrate models addressing the long-term evolution of the nanoscale features,

− to correctly draw conclusions and configure models on the hardening effect of those nanoscale features.

Three sets of data separately reported in the published literature [1-3] have been selected for a critical consideration of the cluster composition in commercial-purity Fe-12at%Cr irradiated at 300°C up to a neutron exposure of 0.6 dpa. The first set of data was derived from the nuclear component of small-angle neutron scattering (SANS) [1]. The second set is based on an atom probe tomography (APT) study [2]. The APT needles were prepared from the bulk of the SANS sample. The third piece of information is adopted from an investigation of the same material by means of positron annihilation spectroscopy (PAS) [3]. The SANS results [1] were found to be consistent with the assumption that the dominant scatterers are α’-phase particles near thermodynamic equilibrium. A composition far from equilibrium was deduced from the APT data [2]. Similar apparent discrepancies were reported in the literature for other systems. In the presentation an effort to overcome the apparent discrepancy will be reported in detail. The basic weakness of SANS is the integrating and one-parametric nature of the composition-related information hidden in the nuclear scattering contrast. Weaknesses of APT are a possible overestimation of Fe in clusters due to trajectory overlap and the insensitivity to vacancies. In the latter respect, PAS data add a value to the comparison. Other factors will be considered as well. The approach is based on the idea that the measured value of the Porod invariant of nuclear SANS can be directly compared with the corresponding quantity calculated solely from the APT data, namely volume fraction and cluster composition. Careful treatment of all potential factors of uncertainty allows the Fe fraction in the clusters to be estimated. References

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[1] F. Bergner, A. Ulbricht, C. Heintze, Estimation of the solubility limit of Cr in Fe at 300 °C from small-angle neutron scattering in neutron-irradiated Fe–Cr alloys, Scripta Materialia 61 (2009) 1060–1063.

[2] V. Kuksenko, C. Pareige, C. Genevois, F. Cuvilly, M. Roussel, P. Pareige, Effect of neutron-irradiation on the microstructure of a Fe–12at.%Cr alloy, Journal of Nuclear Materials 415 (2011) 61–66.

[3] M. Lambrecht, L. Malerba, Positron annihilation spectroscopy on binary Fe–Cr alloys and ferritic/martensitic steels after neutron irradiation, Acta Materialia 59 (2011) 6547–6555.

42

K-15

INTEGRATED MATERIALS LIFETIME ASSESSMENT FOR DEMO: NEUTRON-TRANSPORT, INVENTORY, AND ATOMISTIC

CALCULATIONS

M.R. Gilbert, S.L. Dudarev, S. Zheng, L.W. Packer, and J.-Ch. Sublet

Culham Science Centre, Abingdon OX14 3DB, Oxfordshire, UK Email: mark.gilbert@ccfe.ac.uk

The high-flux neutron-irradiation environment produced by the fusion plasma in a commercial fusion reactor will alter the properties of materials used in the construction of the reactor vessel. As well as producing defect microstructure due to the displacement of atoms from their lattice sites, neutron bombardment also leads to nuclear reactions that alter the chemical composition of the irradiated material. Despite the fact that the chemical elements produced by transmutation reactions are normally from the same region of the periodic table as the original constituent atoms of the material, in sufficient quantities they can still cause major changes to the mechanical and structural properties of the material. Even worse, some of the nuclear reactions also generate gas particles, such as helium, which can have life-limiting consequences for components at even modest concentrations. In this paper we perform neutron-transport calculations on a recent conceptual design for the demonstration fusion reactor (DEMO). The calculations reveal the marked variation in irradiation conditions that would be experienced by materials as a function of position within the same components, but also between different regions, such as the first wall and the divertor. Inventory calculations show that the consequences for the transmutation-induced changes in chemical composition under these different neutron-irradiation environments will be significant. In particular, the levels of gas production, most notably helium, vary widely as a function of depth into the reactor first wall. To explore the consequences associated with helium production in materials, we develop a simple model to estimate the critical bulk helium concentrations that can produce grain-boundary embrittlement. Using the inventory results we then obtain critical lifetime estimates for several of the main candidate elements for fusion applications. These results, which show significant variation of radiation and transmutation effects as a function of both material and reactor position, illustrate, on the one hand, the need for careful consideration of the effect of neutron irradiation on components within the framework of reactor design studies, and on the other show the encouraging progress in the predictive power of computer simulation algorithms that should help select materials for commercial fusion. This work was funded by the RCUK Energy Programme under grant EP/I501045 and the European Communities under the contract of Association between EURATOM and CCFE. This work was also carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

43

K-16

VOID SWELLING AND IRRADIATION CREEP OF FERRITIC-MARTENSITIC ALLOYS AT VERY HIGH DPA LEVELS PRODUCED

BY EITHER NEUTRON OR SELF- ION IRRADIATION

F. A. Garner1*, M. B. Toloczko2, V. N. Voyevodin3, Yu. V. Konobeev4, A. Povstyanko5, B.H. Sencer6, S. A. Maloy7

1Radiation Effects Consulting, Richland WA, USA

2Pacific Northwest National Laboratory, Richland WA, USA 3Kharkov Institute of Physics and Technology, Kharkov, Ukraine

4Institute of Physics and Power Engineering, Obninsk, Russia Federation 5Research Institute of Atomic Reactors, Dimitrovgrad, Russian Federation

6Idaho National Laboratory, Idaho Falls, ID, USA 7Los Alamos National Laboratory, Los Alamos, NM, USA

*Email: frank.garner@dslextreme.com In order to achieve high burn-up of fast reactor fuel it is necessary to contain the fuel in cladding that resists void swelling and irradiation creep more effectively than is achieved using austenitic steels. The current first-generation candidate swelling-resistant alloys are ferritic and ferritic-martensitic steels, with second-generation alloys being oxide dispersion-hardened variants of these steels. It is well known that ferritic alloys as a class swell and creep less than do austenitic alloys. Whereas current maximum fuel burn-ups of 10-11% are attained in fast reactors for doses of ~100-150 dpa using swelling-prone austenitic steels, higher burn-ups require 250-300 dpa, while some other reactor concepts envision doses of 400-600 dpa. The question arises whether acceptable levels of swelling and irradiation creep of ferritic-martensitic steels at such high doses are achievable. A review is presented of recent high-dose irradiation studies on HT9 and EP-450 conducted in FFTF and BOR-60 with maximum doses of 200 and 163 dpa, respectively, and also on EP-450, EP-823 and EP-852 irradiated side-by side in BN-350 to doses of 61 dpa. Additionally data on MA957 are available from FFTF to ~100 dpa. Irradiation creep strains in these ferritic-martensitic steels have been found to be less by a factor of ~2 compared to austenitic steels, but creep strains are sometimes obscured by concurrent precipitation and recovery strains, often of opposite sign. Dispersoids in MA957 were found to reduce thermal creep at higher temperatures but not to reduce irradiation creep to doses of ~100 dpa. To probe swelling at doses >200 dpa ion irradiation is being used. While neutron data at 150-200 dpa indicate that acceleration of swelling rate in EP-450 and HT9 is just beginning, ion bombardment shows that swelling behavior in this alloy class is bilinear (transient and then steady-state) with a post-transient steady-state swelling rate of ~0.2%/dpa, one fifth the rate of austenitic steels. The steady-state ion-induced swelling rate agrees with an earlier prediction arising from simple Fe-Cr binary alloys irradiated in EBR-II and FFTF. The transient regime, however, is long in grains that are composed of decomposed martensite (>300 dpa), but is much shorter in ferrite grains (~150 dpa) even when containing dispersoids. Such divergent behavior is evident in duplex alloys such as EP-450 where both grain types exist. HT9 without ferrite resists the onset of high swelling but eventually reaches ~25% at 600 dpa.

44

K-17

EFFECT OF DISLOCATION LOOP DECORATION ON RADIATION-INDUCED HARDENING IN DIFFERENT IRON ALLOYS

L. Malerba1, D. Terentyev1, G. Bonny1, V. Kuksenko2, C. Pareige2, P.

Pareige2, P. Olsson3, G. Monnet4

1SCK CEN, Boeretang 200 – 2400 Mol, Belgium 2CNRS Rouen, France

3KTH Stockholm, Sweden 4EDF R&D, France

Email: lmalerba@sckcen.be Radiation-induced hardening and embrittlement (RIHE) is the main factor limiting the lifetime of operating nuclear reactors and, even though higher operation temperatures will somewhat reduce its relevance in innovative nuclear systems, it will remain an important issue to tackle there as well. RIHE is caused by the radiation-induced formation of nanostructural features that hinder the movement of dislocations. In reactor pressure vessel (RPV) steels these nanofeatures are diffuse clusters of solute atoms (typically Cu, but also Mn, Ni, Si and P). In particular, Cu-free clusters rich in Mn and Ni (so-called late-blooming phases, LBPs) might be the cause of additional RIHE beyond the dose corresponding to 40 years of operation (~0.1 dpa), thereby questioning the possibility of extending the life of existing reactors beyond their design. Recent microstructural investigations on Fe-Cr alloys containing also small quantities of other impurities have revealed significant density of diffuse clusters very similar to the LBPs observed in RPV steels, with Cr replacing Mn. These features are likely to be found in irradiated ferritic-martensitic steels, too. At the same time, mechanical testing reveals that Fe-Cr alloys harden under irradiation more than iron. In this work it is shown, based on a multiscale modelling approach, that the two families of solute clusters observed in RPV steels and in Fe-Cr alloys are likely to have the same origin, namely segregation of solute atoms on small dislocation loops. It is also shown that, at least in the case of the Fe-Cr alloys, Cr-decorated dislocation loops are stronger obstacles to dislocation motion than undecorated loops and might be the cause of the increased radiation hardening observed in these alloys. It is argued that the potential contribution to embrittlement of RPV steels caused by the so-called LBPs might thus have the same origin. One can conclude that a unified description of RIHE in any ferritic alloy can be achieved on the basis of a relatively simple mechanism involving not only radiation defects but also, importantly, microchemical processes.

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Contributed Papers

(in the order of the programme schedule)

46

47

C-1

UPDATING STAINLESS STEEL 316L(N) PROPERTIES FOR APPLICATION IN FISSION AND FUSION

F. Tavassoli

Commissariat à l’Energie Atomique et aux Energies Alternatives

Email: farhad.tavassoli@cea.fr Austenitic stainless steels have had excellent service records in both light and fast breeder reactors. It was based on this experience that Type 316L(N) was selected for ITER vessel and internals. It was also for the same reason that this steel has been selected for the sodium cooled Gen-IV reactors. It is therefore of no surprise that when EFDA launched a number assessment activities regarding conceptual design activities on fusion DEMO models 1 (a conservative baseline design) and 2 (an optimistic design), they included 316L(N) steel. Under the European Fast Breeder activities, in the 80s and the early 90s, most of the 316L(N) properties were updated and entered in the newer editions of RCC-MR. Thereafter however, these activities were reduced or halted and are only recently picked up under the Gen-IV program. It was indeed the convergence of the fusion national programs into an international collaboration under ITER and selection of 316L(N)-IG for ITER vessel and internals that filled the gap. The data obtained under fusion program allowed updating of the 316L(N) database and introduction of the irradiation effects under the ITER Interim Structural Design Criteria (ISDC). However, ITER operating temperature is below 350°C and hence the updates did not cover the high temperature properties. This paper updates the 316L(N) steel properties, with particular emphasis on high temperature properties. It following the new French RCC-MRx code procedures to establish reference design allowables for fusion DEMO models 1 & 2, and sodium cooled Gen-IV reactors.

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C-2

REDUCED ACTIVATION FERRITIC MARTENSITIC STEEL AND FABRICATION TECHNOLOGIES FOR INDIA’S TEST BLANKET

MODULE (TBM) IN ITER

T.Jayakumar1*, M.D.Mathew1, K.Laha1, Shaju K.Albert1, S.Saroja1, A.K.Bhaduri1 and E.Rajendra Kumar2

1Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research,

Kalpakkam, India 2 Institute for Plasma Research, Gandhinagar, Gujarat, India

*Email: tjk@igcar.gov.in

India is one of the countries associated with the development and testing of Test Blanket Module (TBM) in ITER. Accordingly, India has taken up development of 9Cr-W-Ta Reduced Activation Ferritic Martensitic (RAFM) steel, the structural material chosen for TBM and the technologies required for fabrication of TBM. With an objective of developing an India-specific RAFM steel, four heats of RAFM steel with W and Ta content varying in the range 1 – 2 wt. % and 0.06 -.014 wt. % respectively were metled. The steels were melted through VIM and VAR routes with strict control over the radioactive tramp elements (Mo, Nb, B, Cu, Ni, Al, Co, Ti) and on the elements that promote embrittlement (S, P, As, Sb, Sn, Zr, O). Extensive mechanical testing and metallurgical characterisation of these steels were carried out. The ductile to brittle transition temperature (DBTT) of the steel increased with both tungsten and tantalum content. Tensile strength of the steel was found not to influence significantly with the increase in tungsten content, however decreased marginally with the increase in tantalum content with the consequent increase in ductility. Creep rupture strength of the steel was found to increase significantly with tungsten content whereas it decreased with the increase in tantalum content. Fatigue life of the steel was found to increase with the increase in tungsten and tantalum contents, however extensive cyclic softening was exhibited by the steel with a tungsten content greater than 1.4 wt.%. RAFM steel having 1.4 wt. % tungsten with 0.06 wt. % tantalum tends to have better combination of strength and toughness and is considered as India-specific RAFM steel. Structure-property correlation, microstructural evolution on long term ageing, long-term creep test, low cycle fatigue test and fatigue and creep crack growth studies on this steel are in progress. The joining technologies adopted for the fabrication of TBM are Hot Isostatic Pressing (HIP) to produce first wall of TBM, and technologies of Gas Tungsten Arc (GTA), Electron Beam (EB), Laser and Laser Hybrid welding. Welding consumables for joining this steel has been developed and characterized. Properties of the GTA welds met the entire specification requirement comparable with that of the base metal. This consumable has also been successfully used to carry out Hybrid laser welding of RAFM steel. The procedure for EB welding to join plates of thicknesses up to 12 mm has been developed. Impact tests conducted on EB welds showed that toughness of the weld metal in the as-welded condition is comparable to that of the base metal. A box structure, which simulates one of the components of TBM, has been fabricated using EB to demonstrate the applicability of the process for component fabrication. Laser welding of 6 mm thick plates of RAFM steel has

49

also been carried out successfully and the properties of the weld joints have been found to be satisfactory. Procedure for laser Hybrid welding of 12 mm thick RAFM steel is being developed. Initial trials carried out using stainless steel plates has demonstrated that plates with internal channels, as required for the first wall of TBM can be produced by joining plates with pre machined grooves by Hot Isostatic Pressing. Trials are in progress to develop a procedure for making this TBM component using RAFM steel. The paper will discuss metallurgical characterisation of the Indian-specific RAFM steel and current status of the fabrication technologies being developed for fabrication of Indian TBM to be tested in ITER.

50

C-3

FRACTURE PROPERTIES OF NEUTRON IRRADIATED ODS EUROFER97 STEEL

N.V. Luzginova*, T. Bakker, H. Nolles, P. ten Pierick, M. Jong, M. Kolluri

Nuclear Research and consultancy Group (NRG)

Westerduinweg 3, 1755 ZG Petten, The Netherlands *Email: luzginova@nrg.eu

Eurofer97 steel is a candidate structural material for the fusion power plant. It is well known that ferritic/martensitic steels have satisfactory radiation and swelling resistance compared to the austenitic steels. However, they can suffer from the grain boundary and matrix creep at temperatures above 550 oC. Oxide dispersion strengthened (ODS) steel is expected to have the improved creep resistance and to be used at temperatures as high as 650 °C, which is 100 °C higher than the current operating limit for the conventional ferritic/martensitic steels. The EU reference batch of ODS Eurofer97 steel produced by PLANSEE in different product forms has been distributed by FzK in 2005. A block of 260 x 450 mm machined from the 6 mm plate was delivered to NRG for mechanical characterization and investigation of the irradiation response of this material. The nominal composition of the EU batch of ODS Eurofer97 steel is Fe-0.1C-9Cr-1W-0.2V-0.1Ta-0.3Y2O3 (wt.%). The well-tested irradiation device known as SUMO, an acronym for in-SodiUm steel Mixed specimens IrradiatiOn, has been used to irradiate ODS Eurofer97 steel specimens. The specimen holders named SUMO-11 and SUMO-12 loaded with specimens for mechanical testing have been irradiated in the High Flux Reactor (HFR) core position G3 during operating cycles 4 and 11, respectively. Target irradiation temperatures are 300, 450 and 550 ºC for both specimen holders and maximum target dose levels are 1 and 3 dpa, respectively. It has been shown that the irradiation hardening of ODS Eurofer97 steel occurs at 300 oC, whereas after irradiation at 450 and 550 oC almost no changes in tensile and fracture toughness properties are observed in comparison with unirradiated material. In the present paper the fracture toughness properties of ODS Eurofer97 steel in the transition region are discussed and compared with the values obtained by the different laboratories. Besides, impact properties including Charpy impact energy and lateral expansion results as well as shear fracture area measurements are discussed for different irradiation temperatures and neutron doses, and compared to the impact properties of unirradiated material. This work is performed under the contract of the European Fusion Development Agreement (EFDA, TW5-TTMS-006-D3) with a financial support from the European Commission and the Dutch Ministry of Economic Affairs.

51

C-4 MATERIAL ISSUES FOR DESIGN AND LICENSING OF MYRRHA ADS

SYSTEM

S. Gavrilov*, R. Fernandez, M. Schyns, H. Aït Abderrahim

SCK•CEN, Boeretang 200, B-2400 Mol, Belgium *Email: sgavrilo@sckcen.be

The deployment of innovative nuclear systems including MYRRHA ADS system relies primarily upon the availability of structural materials able to withstand their harsh operation conditions, in particular with respect to the material compatibility with the coolant under intense irradiation. The materials selection and qualification are critical issues for successful development of such nuclear systems. The selection of structural materials should rely on a number of their intrinsic properties with respect, among others, to thermo-mechanical loading conditions, coolant-material interaction effects, irradiation and their synergetic effects of all previously mentioned. Therefore, it is important to be able to appropriately characterize these materials. However, the behaviour of materials under operation conditions of innovative nuclear systems is not appropriately covered by the available testing and evaluation standards requiring updates of existing standard procedures and sometimes development of new ones. Today, individual laboratories rely on their own experience and on experimental and irradiation facilities available to them. Although it is possible to extract helpful qualitative information, it is difficult to interpret the various data quantitatively. The main goal of materials R&D and qualification programme for MYRRHA candidate materials is to provide reliable material property data for design and licensing of MYRRHA. It also aims to assist with the following activities:

− design of various components; − fuel development; − safety analysis; − coolant technology development; − elaboration of the surveillance program.

At the moment the efforts are distributed over the following four overlapping activities:

− Identification of key material issues for design and licensing of MYRRHA; − Development of test and evaluation guidelines for structural materials

characterisation; − Assessment of material properties; − Development of testing infrastructure.

The three major material degradation effects that have been identified so far for nuclear systems with lead-bismuth coolant and for MYRRHA in particular are liquid metal corrosion, environmentally assisted cracking and irradiation effects. In this paper we will report on the material R&D approach for MYRRHA and the progress accomplished so far.

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C-5

DEVELOPMENT OF HIGH TEMPERATURE MATERIALS FOR NUCLEAR APPLICATIONS

D. Riley*, D. Carr, H. Zhu, R. Harrison, L. Edwards

Institute of Materials Engineering, Australian Nuclear Science and Technology

Organisation, Locked Bag 2001, Kirrawee DC, NSW, Australia *Email: dry@ansto.gov.au

Proposed operating environments of the next generations of fission and fusion power reactors presently exceed the properties of most materials. Extremes of temperature, radiation fluence and chemical corrosion rapidly degrade the performance of the current generation of materials. Through a combination of limited radiation tolerance, enhanced plasticity, volumetric distortion and embrittlement, the predictable lifetime of future reactors is largely unknown. Additional economic concerns associated with constructing and decommissioning potential reactor designs still largely persist through a lack of reliable materials data. Broadly considered, two alternate approaches to materials development exist; (i) Enhancement of the current generation of materials, or the (ii) Design and certification of novel materials. The Institute of Materials Engineering (IME, ANSTO) has several programs associated with both the improvement of common nuclear materials (Fe, Zr and Ni based alloys) and the research of novel materials classes (UHTCs and intermetallics). Specific development and enhancement of ODS steels has centred on improving the ductility of these precipitation hardened materials, thereby improving their creep and fatigue resistance. While several years research into the synthesis of mix carbide and oxy-carbide ceramics has improved thermal shock resistance and machinability, thereby lowering the fabrication and maintenance costs of these ultra-high temperature ceramics (UHTCs). Finally, recent investigation of high temperature intermetallics was aimed at assessing the feasibility of alternate materials from both the aerospace and biomedical fields, integrating institutional expertise in grain structure refinement and certification processes. A program overview and novel results will be presented on topics of ODS development and testing, UHTC fabrication and validation as well as radiation tolerance of select intermetallics by high-energy ion-beam irradiation. Progress in the necessary development of these materials for final application at high temperature and high radiation damage levels will also be presented with respect to optimisation of microstructure and morphology.

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C-6

EFFECT OF CRACK LENGTH–TO–WIDTH RATIO ON CRACK RESISTANCE OF HIGH Cr–ODS STEELS

R. Chaouadi*, M. Ramesh and S. Gavrilov

SCK•CEN, Boeretang 200, 2400 Mol, Belgium

*Email: rachid.chaouadi@sckcen.be Oxide dispersion strengthened (ODS) steels with high Cr–content have received a particular attention within the nuclear materials community for application to both advanced fission reactors and fusion systems. In comparison to standard high Cr–steels, the expected operation temperature range can be extended to about 650°C because of their improved creep resistance. However, their crack resistance behavior in the high temperature range was less investigated. The aim of the present paper is to provide some insight on their fracture behavior at high temperature. Crack resistance measurements were performed on compact tension specimens at 650°C on both shallow as well as deep crack by adapting the crack length–to–width ratio. In particular, the shallow crack simulates surface crack that might be important for fuel cladding application.

54

C-7 CRYSTALLOGRAPHIC RELATIONSHIP OF OXIDE PARTICLES WITH

MATRIX IN 12CR ODS STEEL

J. Jang1*, T. K. Kim1, S.H.Kang1, S. S. Kim1, X. Mao1, 2, K.H. Oh2

1Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon 305-353, Republic of Korea

2Department of Materials Science and Engineering, Seoul National University, Seoul 151-742, Republic of Korea

*Email: jjang@kaeri.re.kr The superior strength of ODS steels is not only due to the dispersion of nano-sized oxide particles, but also due to the combination of fine grains and high density dislocations. The stability of grain boundaries and dislocations is thus important to the strength at high temperatures. Nano-sized oxide particles could stabilize the grain boundaries by Zener’s pinning effect and also could act as obstacles to dislocation movement. One of the main factors that affect the pinning effect of oxide particles on grain boundaries and dislocations would be the crystallographic relationship of the oxide particles with the matrix. Coherent particles are supposed to improve the Zener’s pinning effect on grain boundaries, while suspicious on reducing the pinning effect on dislocation movement. The crystallographic relationship of oxide particles with matrix is thus a fundamental issue to be first clarified. In the present study, a 12Cr ODS steel sample with uniformly dispersed YTaO4 particles was fabricated. The crystallography of monoclinic YTaO4 particles and the atomic structure at the particle/ferrite matrix interface have been examined by SAD (selected area diffraction) and HRTEM. The results show the lattice continuity between YTaO4 particles and the matrix phase by the plane of {110} planes of the matrix. The habit plane determined in most cases is the O M(051) //(011) , where the misfit is within 1% of deviation. Crystallographic relationships of O M(051) //(011) ,

O M[715] //[111] ; O M(121) //(110) , O M[210] //[001] ; O M(051) //(011) ,

O M[715] //[011] and O M(051) //(011) , O M[315] //[113] were found. These crystallographic relationships showed a very small amount of misorientation between the {110} close packed planes of the ferrite matrix and those corresponding planes of YTaO4 particles. HRTEM analyses were applied on the particles ranging from 4 to 80 nm in diameter. It shows that the crystallographic relationship (semi-coherent) sustains for the whole size range observed and the finest particles were not necessarily fully coherent with the matrix.

55

C-8

MICROSTRUCTURES AND STRENGTHENING MECHANISMS OF A NANOSTRUCTURED FERRITIC ALLOY AT WIDE TEMPERATURES

J. H. Kim1, T. S. Byun2, D. T. Hoelzer2, S. W. Kim1, J. T. Yeom1, J. K. Hong1

1Special Alloys Group, Korea Institute of Materials Science, Changwon, South

Korea 2Materials Science and Technology Division, Oak Ridge National Laboratory,

Oak Ridge, TN 37831, USA *Email: kjh1754@kims.re.kr

The temperature dependence of strengthening mechanisms in the nanocluster-strengthened 14YWT alloy was investigated to elucidate the relative significance of contributing mechanisms in different temperature ranges and to model the temperature dependence of yield strength. Microstructural and mechanical characterizations were conducted to provide the key deformation microstructure and mechanical data at temperatures ranging from -196 to 1000 °C. Microstructural information before and after tensile deformation, including crystallographic texture, dislocation structures, and nanoclusters, was obtained from the focused ion beam lift-out specimens using EBSD, TEM, and APT (atom probe tomography) techniques. As major strengthening mechanisms, the Peierls stress, grain boundary strengthening, direct nanocluster strengthening, and dislocation forest hardening, were taken into account and their roles and characteristics in different temperature ranges were extensively discussed. The contribution of grain boundary strengthening to overall strengthening was most significant. An yield strength prediction model was established and compared with the experimental data. Both the simple linear and the root mean-square methods for the summation of strengthening components were explored and the later model resulted in excellent agreement with experimental data. A validation of the proposed approach is attempted by applying to the yield strength of other alloys. ______________ Keywords: Ultrafine grained microstructure; Modelling; Yield phenomena; Mechanical alloying; Dislocation structure

56

C-9 EVALUATION OF MECHANICAL PROPERTIES OF ODS ALLOYS IN

THE GETMAT PROJECT

M. Serrano1*, A. García-Junceda1. M.Hernandez1, M. Rieth2, R. Lindau2, G. Müller2, M. Yurechko2, Y. de Carlan3, R. Chaouadi4, M. Hoffman5, C.

Feuillete5, P.Haehner6, J.Chen7, S.Concari8

1CIEMAT, Madrid, Spain 2Institute for Applied Materials, KIT, Eggenstein-Leopoldshafen, Germany

3 Nuclear Materials Department, CEA Saclay, France 4SCK CEN, Mol, Belgium

5 MPA, Stuttgart, Germany 6 IET-JRC, Petten, The Netherlands

7 PSI, Willingen, Switzerland 8 RSE, Piacenza, Italy

*Email: marta.serrano@ciemat.es Iron based oxide dispersion strengthened alloys (Fe-ODS) are promising candidates for structural materials operating at high temperature, high neutron load and corrosive coolants. In particular, they are foreseen as a long term solution for cladding of Generation IV fission reactor due to their high swelling resistance, very good high temperature behaviour and compatibility with liquid metals. In this context, one of the objectives of the European Commission FP7 project GETMAT is to provide reference ODS alloys and evaluate their basic physical and mechanical properties in Gen-IV and Transmutation relevant environments. In this paper an overview of the mechanical characterization, including tensile, impact small punch, creep, fatigue and fracture toughness at high temperature of 12Cr-ODS and 14Cr-ODS alloy performed within GETMAT is presented.

57

C-10

LOW ACTIVATION CRMOV STEEL

F.Gillemot

Centre for Energy Research Hungarian Academy of Sciences, Budapest, Hungary

Email: gillemot.ferenc1@t-online.hu; gillemot@aeki.kfki.hu Since decades low-alloyed steels have been developed and are extensively used as Reactor Pressure Vessel (RPV) materials. In particular, RPV materials with low content of impurities for the second and third generations of Nuclear Power Plants (NPPs), have proven very high stability to radiation damage and their response to radiation is very well understood as well as the effect of deleterious impurities. The industry has experience in production, machining, welding etc. of these steels. Some components of the future nuclear power reactors will be operated at relatively high temperatures and exposed to higher neutron doses requires the development of new alloys operated at high temperature and tough against high neutron dose. Beside this special steels, it is also necessary to have low activation cheap structural materials for all structures, where the requirements exceed the present LWR reactor properties, but the application of the future ODS and ferritic martensitic highly creep resistance steels is not necessary (e.g. RPV of the SCRW reactor, or components of the planned ALLEGRO reactor etc.). The activation of the CrMoV low alloyed steels with low nickel or free of nickel is lower than any other RPV steel used presently and most of the alloying components decay time is short. After some years of shut down the activity level is low enough for decommission and after 40-100 years decay time it may be reused. Industrial experience in producing and operating the CrMoV low alloy steels in fossil plants up to 560 °C has been gained during the last 50 years.. It shows that this type alloy is not sensitive for thermal ageing. The tensile and creep properties, weldability of it is also well known. Irradiation data of the WWER-440 reactors (leading in EOL fluence among the PWR-s, and having accelerated surveillance results exposed for 15-25 years of irradiation and up to 1.5*1021 n/cm2 E>1 MeV) verifies the radiation stability of this type of steel. In the irradiation temperature range of 400 - 500 ºC the CrMoV steel radiation embrittlement is expected to be very small, but few data is available. Application for future reactors of this steel requires only studying the long term irradiation embrittlement on high temperature, including irradiation assisted creep and low cycle fatigue. The presentation will summarize the existing information and propose further research

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C-11

INVESTIGATION ON DIFFERENT OXIDES AS CANDIDATES FOR NANO-SIZED ODS

PARTICLES IN REDUCED-ACTIVATION FERRITIC (RAF) STEELS

J. Hoffmann*, M. Rieth, R. Lindau, M. Klimenkov, A. Möslang,

Institute for Applied Materials (IAM-AWP), Karlsruhe Institute of Technology (KIT), Karlsruhe, Germany *Email: j.hoffmann@kit.edu

Future generation reactor concepts are based on materials which can stand higher temperatures and higher neutron doses in corrosive environments. 9 to 14 wt% ODS steels – produced by mechanical alloying - are typical candidate materials for future structural materials in nuclear power plants. Yttrium has proven to be an excellent addition to ferritic steels during mechanical alloying to form nano-sized dispersion oxide particles during compacting of the material. These oxide particles have various beneficial effects on the material such as improved high temperature properties and creep strength combined with the excellent corrosion resistance of high-chromium ferritic steels. However, there might be potential for improvements by choosing different oxides. In this present work, four different oxides (Magnesium-, Lanthanum, Cerium- and Zirconium-oxide) are selected by looking at thermal stabilities and Gibbs-Free-Enthalpies of various chemical compositions. These oxides are mixed and mechanically alloyed with ferritic steel powder (Fe13Cr1W0.3Ti) and compared to a reference material produced with an addition of yttrium-oxide (Fe13Cr1W0.3Ti + Y2O3). Compacting is done by hot-isostatic pressing (HIP), which is followed by hot-rolling and annealing. The materials are then characterized with mechanical tests and detailed microstructural investigations by transmission electron microscopy with Energy-Dispersive X-ray spectroscopy (EDX). All further results of the mechanical testing and microstructural characterizations are analyzed, compared, and discussed in this paper.

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C-12

IRRADIATION EMBRITTLEMENT CHARACTERIZATION OF THE EUROFER 97 MATERIAL

R.Kopriva* and M.Kytka

Integrity and Technical Engineering Division

Nuclear Research Institute Rez, Czech Republic *Email: kor@ujv.cz

The paper summarizes results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV. Paper is mainly focused on the comparison of results from sub-sized specimens with the standard Charpy type specimens. EUROFER97 is the European candidate RAFM steel for use as a structural material in GEN IV reactors and fusion energy systems. High chromium ferritic/martensitic steels were developed more than 50 years ago and have been used for a long time in the power-generation industry as boiler and turbine materials, as well as for other applications. In the early 1970’s, they were considered for fast breeder fission reactors and later for fusion applications. In the European Union, within the long term programme of EFDA (European Fusion Development Agreement), remarkable effort has been spent by several institutes for the characterization and optimization of a reference EUROFER 97 steel. The task designated “Irradiation Performance of EUROFER” (TTMS-001) is one of the most important within the Long Term Programme and involves numerous European research institutes. The programme in the EU was progressing with irradiations of EUROFER 97 up to a wide range of radiation damage: from 0.3 to 1.0, 3 and 5, 10, 15, 30 up to 70–80 dpa. The investigation of irradiation performance limits of EUROFER 97 included the irradiation of various product forms at different temperatures, and post-irradiation examinations (both mechanical tests and microstructural investigations). Nuclear Research Institute Rez have participated in the EFDA programme. The static and dynamic fracture toughness properties of EUROFER 97 base material and TIG weld metal were acquired in initial and irradiated state. The irradiation was carried out in Chouca rig in the core of LVR-15 research reactor (NRI Rez) and the samples were irradiated to approx. 2.5 dpa at max. 235 °C. Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region – static fracture toughness and dynamic fracture toughness properties, with sub-size three point bend specimens (dimensions 27 x 4 x 3 mm) and standard Charpy type specimens. Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness KJC and KJd data were also evaluated with the “Master curve” approach. Moreover, J-R dependencies were determined and analyzed. The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given. Results from irradiation of EUROFER 97 show that this steel – base metal as well as weld metal – is suitable as a structural material for reactor pressure vessels of innovative nuclear systems – fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of Eurofer 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that

60

there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with Eurofer 97 base metal.

61

C-13 FEASIBILITY OF JOINING OF 14YWT AND F82H BY FRICTION STIR

WELDING

D.T. Hoelzer*, M.A. Sokolov, K.A. Unocic and Z. Feng

Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, 37831-6136, USA

*Email: hoelzerd@ornl.gov The advanced Oxide Dispersion Strengthened (ODS) 14YWT ferritic alloy was developed for increasing the tolerance to high dose radiation damage, including the accumulation of He, which will help improve the performance and safety of components for advanced nuclear energy systems. The ODS 14YWT ferritic alloy contains a very high internal interfacial area associated with the high concentration of nano-size Ti, Y- and O-enriched clusters, or nanoclusters (NC), and ultra-fine size grains that provide the high sink strength for attracting point defects and He during irradiation. In addition, the ultra-fine size grains and high concentration of NC provide the unique combination of high-temperature strength and low fracture toughness ductile-to-brittle transition temperature (~150ºC) characteristic of 14YWT. One of the major drawbacks that historically has prevented ODS alloys from gaining acceptance in engineering applications has been the problem associated with joining. However, one of the most promising joining methods receiving considerable attention is friction stir welding (FSW) since it is based on solid state joining principle. The purpose of this study is to determine the effectiveness of joining 14YWT and also joining 14YWT with the F82H tempered martensitic steel (TMS) by FSW. The first FSW experiment that was conducted resulted in the successful bonding of 14YWT specimens together and to the F82H plate. The overall consolidation of the sample was good and the interfaces associated with joints between the thermomechanically affected zone (TMAZ) and heat affected zone (HAZ) of 14YWT and between TMAZ of 14YWT and F82H were relatively narrow in width. However, defects were produced by FSW that consisted of a few large pores and a high number density of small pores in the microstructure located on the advancing side of the TMAZ of 14YWT and numerous small pores located at the interface between the TMAZ of 14YWT and F82H. Vickers Hardness measurements across the interfaces showed a decrease from 499 to 376 VHN (~20%) between the HAZ and TMAZ of 14YWT and an increase from 221 to 443 VHN (~100%) between the HAZ and TMAZ of F82H. In both cases, modifications to the grain size of the microstructures in the TMAZ of 14YWT and F82H by FSW may account for most of the changes in the hardness results. TEM specimens were prepared from strategic locations in the weld zone using the lift-out and Focused Ion Beam (FIB) method and investigated to examine the effects of FSW conditions on the stability of the NC dispersion in 14YWT. This presentation will show highlights of the microstructural investigation of specimens from different locations in the weld zones between specimens of 14YWT and between 14YWT and F82H. These results will add to the growing knowledge base that will help determine the feasibility of joining advanced ODS ferritic alloys by FSW.

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C-14

STUDY ON ELECTRON BEAM WELDING AND PWHT FOR CLAM STEEL

Q. Wu1, S. Zheng1, S. Liu1, Ch. Li2*, Q.Huang3

1Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences,

Hefei, Anhui, China 2Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, China

3School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, China

*Email: lcj@ipp.ac.cn The mechanical and metallographical properties of the electron beam weld joint of CLAM steel and the effect of post-weld heat treatment (PWHT) on these properties were investigated. The results showed that the fusion zones (FZ) and the heat-affected zones (HAZ) were typically composed of martensite laths. Much higher microhardness, i.e. 400HV, was observed in the region of FZ and the HAZ, and it was about twice of that of the base material (BM) (~220HV). Although temper heat treatment can decrease the microhardness, but it also decreased the tensile strength and there were no obviously increasement of the impact absorbing energy (Akv). When the joints were performed with quenching at 980。C and then tempering at 720�-760�, the microhardness of the FZ and the HAZ, the tensile strength and the Akv of the joints closed to that of BM. And the optimistic post-weld heat treatment procedure was that quenched at 980� and then tempered at 760�. As a result, twice heat treatments are necessary in order to improve the performance of the weld joints. ______________ Key words: CLAM steel; Electron beam welding; PWHT; Microstructure; Mechanical properties.

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C-15

MICROSTRUCTURE AND MECHANICAL PROPERTIES OF OPTIMISED CUCRZR ALLOY AFTER HIP BONDING CYCLE

P.E. Frayssines1*, J.M. Gentzbittel1, A. Guilloud1, P. Primaux2, T. Soreaux2,

N.François2, S.Heikkinen3, F. Zachia3

1CEA –LITEN-DTBH, Grenoble, France 2Le Bronze Industriel, Suippes, France 3Fusion for Energy, Barcelona, Spain *Email: pierre-eric.frayssines@cea.fr

ITER First Wall (FW) panels are a layered structure made of the 3 following materials: 316L(N) austenitic stainless steel, CuCrZr alloy and beryllium. Two Hot Isostatic pressing (HIP) cycles are included in the reference fabrication route to bond these materials together for the Normal heat Flux (NHF) design supplied by the European Union (EU). To improve this fabrication route, several studies have been done in the past. These studies have demonstrated the possibility to fabricate mock-ups with dissimilar CuCrZr/316LN joint and similar CuCrZr/CuCrZr and 316LN/316LN joints in one HIP cycle at 1040°C. These joining conditions ensure acceptable mechanical properties of the three types of joints without reducing significantly the mechanical properties of 316L(N). To obtain good mechanical properties for the CuCrZr, these joints are then submitted to a solution annealing treatment at 980°C followed by a rapid gas quench. The last step of the manufacturing is a HIP cycle at 580°C, which is done to over-age the CuCrZr alloy and, especially, to join a Be armour layer to the panel. This reference fabrication route ensures sufficiently good mechanical properties to the materials and joints, which fulfil the ITER mechanical specifications, but results often in a coarse grain size for the CuCrZr alloy, which is not favourable, in particular, for the thermal creep properties of the FW panels. To limit the abnormal grain growth of CuCrZr and make the ITER FW fabrication route more reliable, a study has started in 2010 in the EU in the frame of an ITER task agreement. Two material fabrication approaches have been investigated. The first one was dedicated to the fabrication of wrought CuCrZr alloy in close collaboration with an Industrial copper alloys manufacturer. Several CuCrZr alloy manufacturing processes (forging or extrusion) and chemical concentrations (chromium, zirconium) have been investigated. The results show that the combination of one manufacturing process and narrowed chemical concentration range, close to the ITER specification, enable the fabrication of a CuCrZr alloy with limited and predictable grain growth after the reference fabrication route. The second approach investigated was the manufacturing of CuCrZr alloy using powder metallurgy (PM) route and HIP consolidation. This paper presents the main mechanical and microstructural results associated with the two CuCrZr approaches mentioned above. The mechanical properties of wrought CuCrZr/wrought CuCrZr, wrought CuCrZr/PM CuCrZr, wrought CuCrZr/316L(N) and PM CuCrZr/316L(N) mock-ups fabricated using the reference route are also presented.

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C-16 FAIL SAFE AND COST EFFECTIVE FABRICATION OF A FIRST WALL

BY DIFFUSION WELDING

L. Commin1*, M. Rieth1, B. Dafferner1, H. Zimmermann1, D. Bölich1, S. Baumgärtner1, R. Ziegler1, S. Dichiser2, T. Fabry2, S. Fischer2, W.

Hildebrand2, O. Palussek2, H.Ritz2, A. Sponda2

1Institute of Applied Materials (IAM-AWP), Karlsruhe Institute of Technology, Germany

2Technische Infrastruktur und Dienste (TID), Karlsruhe Institute of Technology, Germany

*Email: lorelei.commin@kit.edu Helium Cooled Lithium Lead and Helium Cooled Pebble Bed concepts have been selected as European Test Blanket Modules (TBM) for ITER. The TBM fabrication will need the assembly of six Reduced Activation Ferritic Martensitic steel sub-components, namely First Wall, Caps, Stiffening Grid, Breeding Units, Back Plates/Manifolds, and Attachment system. The fabrication of the First Wall requires the production of cooling channels inside 30 mm thick bended plates. For this specific component, the main issues consist of the lack of accessibility of some areas to join, the process tolerances, the dimensional stability and the resulting assembly mechanical properties. Several fabrication routes have been already investigated, which involve diffusion welding and fusion welding (electron beam, laser beam, hybrid MIG/laser). In this study, an alternative processing method was developed, based on Hot Isostatic Pressing of inner pipes within two half-shells. This method presents some major advantages over the existing ones, in particular its inherent fail-safe design due to the application of the double containment principle, the solely use of cost effective standard fabrication processes and the resulting component dimensional stability. A four channel mock-up was fabricated and analysed to validate the fabrication procedure. The joint quality was assessed using microstructural characterization and Charpy tests. The results show no remaining diffusion weld line and the achievement of suitable mechanical properties. Therefore, this fabrication procedure is fully efficient for the TBM First Wall assembly.

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C-17

INVESTIGATIONS OF RADIATION DAMAGE INFLUENCE ON BEHAVIOR OF STRUCTURAL AND PLASMA FACING MATERIALS

FOR FUSION REACTORS

A.I. Ryazanov*, V.S. Koidan, B.I. Khripunov, S.T. Latushkin, V.B. Petrov, E.V. Semenov, V.N. Unezhev

NRC Kurchatov Institute, Kurchatov Sq. 1, Moscow, 123182, Russia.

*Email: ryazanoff@comail.ru This paper presents a summary of scientific results obtained during the last few years in NRC “Kurchatov Institute” for irradiated structural and plasma facing fusion materials. The main aims of theses researches are the theoretical modelling of point defect accumulations, cascades and sub-cascade formation in fusion structural materials for different irradiation conditions and development of a new experimental method that can be applied for the investigations of a high level of radiation damage influence on a plasma interaction with irradiated and non-irradiated materials. So the developed theoretical model allows determining the total numbers, displacement cross-sections and generation rates for point defect production, cascade and sub-cascade formations for the different fusion neutron energy spectra taking into account elastic and inelastic processes. On the basis of this developed model the numerical calculations for the main important characteristics of radiation damage production including point defect, cascade and sub-cascade formations in different fusion structural materials such as Fe, V, C, Al, Be and W were performed using the neutron energy spectra for fusion reactors: ITER and DEMO. For the comparison of difference in radiation damage production in fusion, fission reactors and under fast charged particle irradiations the additional numerical calculations have been made for neutron energy spectrum of fast atomic reactor HFIR and for fast charged particle irradiations on NRC KI Cyclotron taking into account also elastic and inelastic processes. The detail comparisons of obtained numerical results for radiation damage production in same materials for fusion, fission reactors and charged particle irradiations on cyclotron have been performed with the consideration of elastic and inelastic processes. The developed experimental method at NRC KI allows to investigating of a high level of radiation damage influence on radiation-induced deformation (radiation swelling) in irradiated materials, plasma erosion effects in irradiated plasma facing materials (graphite materials and tungsten) at different temperatures and hydrogen isotope accumulation in these materials relevant to fusion reactor conditions (ITER). A high level of radiation damage in these materials (0.1 - 10 dpa) was achieved by irradiating them with fast charged particles (helium and carbon ions) at the NRC KI Cyclotron simulating fast neutron irradiation in fusion reactor. The plasma erosion effects in irradiated and non-irradiated materials were compared using the linear plasma simulator LENTA at the NRC KI. The performed experimental investigations have shown that the erosion factor of irradiated graphite materials is increased and the tungsten surface structure is changed due to the accumulation of radiation damage. The radiation swelling in tungsten irradiated with 4 MeV helium ions was observed. No influence of irradiation on tungsten erosion rate was observed in these experimental tests. The

66

accumulation of hydrogen isotopes and helium concentrations were measured in the irradiated tungsten. The performed work and the obtained theoretical and experimental results suggest a new promising experimental method for the experimental investigations of plasma effects on fusion structural and plasma facing materials at different irradiation temperatures and different radiation damage levels.

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C-18 DEUTERIUM PERMEATION THROUGH ERBIUM OXIDE COATING ON

RAFM STEELS BY DIP-COATING TECHNIQUE

T. Chikada1*, S. Naito1, A. Suzuki1, T. Terai1, T. Tanaka2, T. Muroga2

1School of Engineering, the University of Tokyo, Tokyo, Japan 2Department of Helical Plasma Research, National Institute for Fusion Science,

Gifu, Japan *Email: chikada@nuclear.jp

Tritium permeation through structural materials is one of critical issues in fusion reactor blankets from the viewpoint of fuel loss and radiological hazard. A promising solution is to fabricate a thin ceramic coating on an inner wall of a metal duct as a tritium permeation barrier (TPB). Recently erbium oxide (Er2O3) coatings have been investigated by various methods and showed high permeation reduction factors (PRFs) by physical vapor deposition. Development of practical technique for large-scale fabrication and formation of the coating on complex-shape substrates has also started by liquid phase methods and metal organic chemical vapor deposition. However, a high TPB performance has not been obtained, indicating fabrication parameters have not been optimized yet. In this study, Er2O3 coatings have been fabricated on reduced activation ferritic/martensitic (RAFM) steels by dip-coating technique, and deuterium permeation experiments have been performed in order to discuss permeation mechanism in the coatings. Dip-coated Er2O3 coatings were prepared by metal-organic decomposition (MOD) method on RAFM steel F82H and JLF-1 plates. The substrate was dipped and withdrawn at a constant speed of 0.5-1.5 mm/s and dried in a furnace for 10 min at 393 K. Then the sample was baked for 10 min at 973 K in a mild oxidation condition, in hydrogen atmosphere with 0.6 % moisture, in order to crystallize Er2O3 without oxidizing the RAFM substrate. The coatings withdrawn at 1.0-1.2 mm/s were uniformly fabricated, while those withdrawn at more than 1.4 mm/s were inhomogeneous after baking process. In the case of the samples withdrawn at less than 0.8 mm/s, the substrates were oxidized because the thickness of the coating was too thin. The thickness of the coating withdrawn at 1.0 mm/s was about 150 nm on each side of the substrate, and the thickness of the oxide layer of the substrate was less than 10 nm by cross-sectional observation. The formation of the cubic-phase of Er2O3 was confirmed using X-ray diffraction. Deuterium permeation measurements were performed by a gas-driven permeation setup at up to 973 K. The sample with the inhomogeneous surface resulted in low PRF values (less than 10). Besides the permeation flux was proportional to the driving pressure, indicating the permeation rate was limited by surface reactions such as absorption and desorption. That means the inhomogeneous surface included cracks and pores in the coating and had large surface area. On the other hand, the sample with uniform coatings showed a reduction factor of more than 103. The permeation flux was proportional to the one-half power of the driving pressure. That indicates the permeation was limited by diffusion of deuterium atoms, and had a small contribution of surface reactions. Therefore, the dip-coated Er2O3 coating by MOD method is a promising technique for an application to DEMO reactors, though it is essential

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to precisely adjust fabrication parameters in order to control the coating microstructure.

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C-19

FRACTURE TOUGHNESS OF NANOSTRUCTURED FERRITIC ALLOYS FOR NUCLEAR APPLICATIONS

M.A. Sokolov*, D.T. Hoelzer and L.Tan

Oak Ridge National Laboratory, Oak Ridge, TN 37831-6151, USA,

*Email: sokolovm@ornl.gov Recent advances in processing and understanding of the oxide-dispersion strengthening mechanisms have resulted in the creation of a new type of materials, namely, nanostructured ferritic (NSF) steels. Elevated temperature strength in these steels is obtained by a high number density of ultra-fine, nanometer-scale complex Ti-Y-O particles dispersed in a ferritic matrix. However it was anticipated that improvements in strength and creep properties of these NSF steels should come in price of low fracture toughness. In this study, fracture toughness of the NSF alloys (12YWT and 14YWT) is compared to fracture toughness of conventional ferritic-martensitic steels like ASTM Grade 92, F82H and EUROFER and conventional ODS steels like EUROFER ODS and MA967. Fracture toughness characterization was performed in the transition region as well as in the ductile region at elevated temperatures (300C and above). The limited data of fracture toughness of these NSF alloys after irradiation in HFIR reactor at 300C will be compared to shifts of fracture toughness of conventional ferritic-martensitic steels like F82H and EUROFER at comparable doses.

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C-20

IRRADIATION BEHAVIOR EVALUATION OF OXIDE DISPERSION STRENGTHENED FERRITIC STEEL CLADDING TUBES IRRADIATED

IN JOYO

S. Yamashita*, Y. Yano, S. Ohtsuka, T. Yoshitake, T. Nagamine, T. Kaito, S.Koyama, K.Tanaka

Oarai Research and Development Center, Japan Atomic Energy Agency

4002, Narita-cho, Oarai-machi, Ibaraki, 311-1393, Japan *Email: yamashita.shinichiro@jaea.go.jp

Oxide dispersion strengthened (ODS) ferritic steels have high resistance to radiation damage and superior long-term creep rupture strengths at elevated temperature, offering a promise of high performance core materials for advanced fission reactor systems, especially for fast reactor system, as well as blanket materials for advanced fusion reactor one. Two types of ODS ferritic steel cladding tubes with different basic compositions and different matrix phases were designed and developed by Japan Atomic Energy Agency (JAEA), and irradiated in the experimental fast reactor JOYO to examine their irradiation performance. In this work, the irradiation behavior of ODS ferritic steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS ferritic steel in addition to the previously accumulated knowledge on ODS ferritic steels. Chemical compositions of the ODS ferritic steel cladding tubes examined were Fe-0.13C-8.84Cr-1.97W-0.20Ti-0.34Y2O3 (9Cr-ODS) and Fe-0.04C-11.34Cr-1.89W-0.25Ti-0.23Y2O3 (12Cr-ODS). Material irradiation test for these ODS ferritic steel cladding tubes were conducted at 683-1108 K to fast fluences ranging from 3.2 to 6.6 x 1026 n/m2 (E>0.1 MeV) in JOYO. The irradiation data concerning hardness, ring tensile property, density and microstructure were obtained through their post irradiation examinations. The results of hardness measurement showed that there was an apparent irradiation temperature dependence on hardness for 9Cr-ODS whereas no distinct dependence for 12Cr-ODS. The post-irradiation ring tensile tests indicated that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 6.6 x 1026 n/m2 (E>0.1 MeV) were decreased by about 20%. Density of the tube specimens before and after irradiation was measured by an immersion method with water, indicating that no significant swelling occurred for all the irradiated specimens. Transmission electron microscope (TEM) observations showed that the radiation-induced defect cluster formation during neutron irradiation was suppressed. It was highly possible due to high density defect sink site such as initially-existed dislocation introduced during tube fabrication process, interface between precipitates including oxide and each matrix. In addition, it revealed that oxide particles were stable up to the maximum doses of this irradiation test from the analyses of TEM micrographs. Details on neutron irradiation behavior of two types of ODS ferritic steel will be presented and discussed in the upcoming meeting.

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C-21

TOMOGRAPHIC ATOM PROBE STUDY OF UN- AND IRRADIATED ODS EUROFER STEEL

S. Rogozhkin1*, A. Aleev1, M. Klimenkov2, R. Lindau2, A. Moeslang2, A.

Nikitin1, P.Vladimirov2, A. Zaluzhnyi1

1SSC RF Institute for Theoretical and Experimental Physics, Moscow, Russia 2Karlsruhe Institute of Technology, Institute for Applied Materials – Applied

Materials Physics (IAM-AWP), Karlsruhe, Germany *Email: Sergey.Rogozhkin@itep.ru

Oxide dispersion strengthened steels possess better high-temperature creep and radiation resistance than conventionally produced ferritic/martensitic steels. This behavior is mainly caused by the presence of highly dispersed and extremely stable oxide particles with diameters of a few nanometers. In this work the nanostructure of European steel ODS Eurofer (9%-CrWVTa) in unirradiated as well as in irradiated up to 32 dpa were investigated. The irradiation was performed in the research reactor BOR-60 at SSC RF RIAR (Dimitrovgrad, Russia) and in heavy ion accelerator TIPr at SSC RF ITEP (Moscow, Russia). Nanoscaled clusters of typically 2 nm diameter containing not only yttrium and oxygen but also vanadium and nitrogen were found in unirradiated state. Moreover, concentration of vanadium in particles was found to be higher than that of yttrium, which indicates the importance of these elements in cluster formation. The estimated average cluster number density is about 2×1024 m-3. A high number density 2÷4×1024 m-3 of ultra fine 1-3 nm diameter nanoclusters enriched in yttrium, oxygen, manganese and chromium was observed in the as-irradiated state. It was observed that after neutron and ion irradiation vanadium atoms had left the clusters, moving from the core into solid solution. The concentrations of yttrium and oxygen in the matrix, as it was detected, increase several times under irradiation. This effect can be tentatively explained by dissolution of the larger yttrium oxide particles (more than 10 nm in diameter) which can be hardly detected by tomographic atom probe, but commonly observed with TEM.

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C-22

MINIATURISED MATERIAL TESTING DEVICES FOR SUPER CRITICAL WATER (SCW) ENVIRONMENT

P. Moilanen1,2, R. Novotny1*, P. Hähner1, P. Janik1,3

1European Commission, JRC-IE, Institute for Energy, Petten, Netherlands

2VTT Technical Research Centre of Finland, Espoo, Finland 3Institute of Chemical Technology, Prague, Czech Republic

*Email: Radek.Novotny@ec.europa.eu A detailed investigation of detrimental processes such as corrosion and stress corrosion cracking (SCC), ageing embrittlement, creep and fatigue etc., which may significantly affect the materials performance in specific water chemistry environments, is of fundamental importance for the development of the Supercritical Water (SCW) reactor concept. In particular, SCC crack growth rate testing is difficult to perform due to complex requirements for the experimental facilities which include water chemistry loops equipped with relevant pH, conductivity, dissolved O2 and H2 sensors, autoclaves with mechanical loading devices, incorporating crack growth rate measurement by direct current potential drop (DCPD) and displacement measurement by linear variable differential transducers (LVDT). All these requirements significantly advance the total materials qualification costs. JRC in cooperation with VTT has been working on a new type of loading devices which are expected to decrease these costs and at the same time guarantee enough reliability and flexibility for both SCC and future irradiation assisted stress corrosion cracking (IASCC) testing to be performed in SCW environments. This paper summarizes the development of a miniature autoclave based material testing system for SCW environment. A prototype double bellows (DB) based pneumatic loading unit which is capable of working under SCW environment has been designed, constructed and tested. Preliminary calibrations for the pneumatic DB loading unit have been performed. The pneumatic loading unit has given reliable results at temperatures of 288 ºC and 550 ºC and 250 bar environmental pressure. The joining and welding procedures of the main parts from Inconel 625 and Nimonic 80A materials have been investigated and performed. Feed throughs for the miniature autoclave have been qualification tested. A pressure compensation system designed to take into account the effect of autoclave pressure on the loading of SCC specimens worked very accurately and reliably during a long term test. Due to this continuous compensation of the autoclave pressure, test load variations caused by pressure variations can effectively be avoided.

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C-23

EFFECT OF LIQUID METALS (Li, Pb) ON THE MECHANICAL AND CORROSION PROPERTIES OF LOW ACTIVATION MATERIALS (V-

4Ti-4Cr ALLOY, ODS STEELS) AS APPLIED FOR FUSION AND FISSION REACTOR CONCEPTS

V. Tsisar1,2*, T. Nagasaka1, T. Muroga1, O. Yeliseyeva2, Z. Zhou3

1National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan

2Physical-Mechanical Institute of NASU, 5 Naukova St., 79601 Lviv, Ukraine 3School of Materials Science and Engineering, University of Science and

Technology Beijing, Beijing 100083, China *Email: valentyn_tsisar@ipm.lviv.ua

“Solid metal – Liquid metal” system is a key subassembly for further fusion and fission reactors. “V-4Ti-4Cr alloy – liquid Li” system is one of the most attractive blanket concept, while “ODS steel – liquid Pb (Pb-Bi)” system is considered as a promising among Generation IV fission reactor concepts. The use of the structural materials facing liquid metals at high temperatures (≥600°C) requires assessment of corrosion aspects of interaction. The character of interaction in “solid metal –liquid metal” system strongly depends on the melt purity with respect to the non-metallic impurities. In this work the effect of liquid Li contaminated with N on the post-test mechanical properties of electron beam welds of V-4Ti-4Cr alloy is evaluated and features of corrosion/oxidation interaction of oxide dispersion strengthened (ODS) steels with liquid Pb depending on the O content in melt are elucidated. “V-4Ti-4Cr alloy – static liquid Li” system. The 4 mm thick plate of V-4Ti-4Cr alloy (NIFS-HEAT-2) was bead-on-plate welded by electron beam (1.5 kW) in high vacuum. The static corrosion tests were carried out for coupon and 1/3 CVN samples placed in V-5Ti capsule filled with Li (8 g) at 700°C for 500 h in glove box under the flowing He continuously purified with respect to O (4-10 wppm). Liquid Li was contaminated by N during test. As tested, samples were cleaned in 30%H2O2 at 5°C to avoid hydrogenation. Nitrogen absorbed by V-alloy from Li resulted in formation of hardened zone (~60µm) in the near-surface layer. Absorbed energy of weld metal measured at 77 K decreased from 10.43 J (as-welded) to 2.25 J after test in Li. Character of fracture changed from ductile (as-welded) to brittle after exposure to Li, probably due to both N absorbed from Li and aging taking place in weld metal at 700°C. “ODS steel – static liquid Pb” system. Samples of reduced activation ferritic-martencitic (Fe-9Cr-2W) and ferritic (Fe-14Cr-2W) ODS steels supplied by University of Science and Technology Beijing (USTB) were tested in stagnant Pb melt at 650°C for 1000h depending on the O content in liquid metal. Samples exposed to pure Pb melt underwent severe corrosion losses averaged 68 and 92 g/m2 for Fe-9Cr-2W and Fe-14Cr-2W steels respectively. Corrosion losses were accompanied by substantial morphological and compositional transformations in the near-surface layers resulted in the formation of dimple-type surface due to preferential corrosion attack (dissolution of Cr and Fe followed by the Pb penetration into steel matrix) proceeded along grain boundaries. Contrary to pure melt, in oxygen containing Pb melt (CO[Pb]≈10-6 wt%) both grades of ODS steels were covered by the thin (≤ 3 µm) Cr-based oxide film. It

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was concluded that fine grained structure of ODS steels intensified corrosion in pure Pb melt in comparison with conventional (coarse-grained) steels. In oxygen-controlled melt, in contrast, fine-grained structure plays positive role providing formation of Cr-rich oxide film instead of duplex magnetite scale typical for steel`s oxidation in heavy metal melts (Pb, Pb-Bi).

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C-24

EVALUATION REACTOR PRESSURE VESSEL STEELS BY POSITRON ANNIHILATION POINT OF VIEW

V. Slugeň1*, H. Hein2, S. Sojak1, J. Veterníková1, M. Petriska1, V. Sabelová1,

M. Pavúk1, R. Hinca1

1Institute for Nuclear and Physical Engineering, Slovak University of Technology,

Ilkovičova 3, 81219 Bratislava, Slovakia 2AREVA NP GmbH, Paul Gossen Strasse 100, 91 001 Erlangen, Germany

*Email: vladimir.slugen@stuba.sk This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generations of Russian RPV steels seems to be fully comparable with German steels and their quality enables prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steel is very low due to the dynamic recovery of radiation-induced defects at reactor operating temperatures. PAS techniques can be applied effectively also for evaluation of microstructural changes caused by extreme external loads simulating irradiation by proton implantation and for the evaluation of the effectiveness of post-irradiation thermal treatments. Therefore, we would like to use our results collected during last 20 years from measurement of different RPV-steels in “as received”, irradiated and post-irradiation annealed and compare them to results where the real neutron irradiation will be replaced by proton implantation.

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C-25 EVOLUTION OF THE MICROSTRUCTURE OF FECR ALLOYS UNDER

NEUTRON AND ION IRRADIATION

M. Hernández-Mayoral1*, C. Pareige2, V. Kuksenko2, P. Pareige2, P. Desgardin2, B. Décamps2, F. Bergner3, A. Ulbricht3, C. Heintze3, L. Malerba4, M. Lambrecht4, A. Idhil5, C. Borca5, M. Samaras5, A. Nordlund6, P. Cartemo6

1CIEMAT, Madrid, Spain;

2CNRS, France; 3HZDR, Rozendorf, Germany;

4SCK•CEN, Mol, Belgium; 5PSI, Willingen, Switzerland;

6University Chalmers, Sweden *Email: m.mayoral@ciemat.es

High Cr (9-14%Cr) ferritic/martensitic steels are considered as one of the most promising candidates for structural materials for Generation IV and fusion reactors. Therefore, the understanding of the evolution of properties under operation conditions is of primary importance. In the present work we report on the investigation carried out within the GETMAT Project, funded by the 7th Framework Program of the EU. The aim of the work is to obtain a complete description of the nanostructure and its evolution in FeCr alloys under irradiation, taking into account that different variables will affect the microstructure and, hence, the behaviour of materials under irradiation. This information will be suitable for the further validation of simulation methods which are being developed in parallel to the experimental work within the project. The materials considered in the study were four commercial purity FeCr binary alloys with varying Cr content and the steel T91 (9%Cr). Two types of irradiation experiments were undertaken. On the one hand, the materials were neutron irradiated at 300ºC in the test reactor BR2 at SCK-CEN (Belgium) up to 0.06, ~0.6 and ~1 dpa. On the other hand, the same set of materials were ion irradiated at different temperatures (100, 300 and 420ºC) and doses (1 and 5 dpa) at the Ion Beam Centre at HZDR (Germany). Different advanced materials characterisation techniques are applied: TEM, SANS, APT, PAS and EXAFS. The complementarity of the techniques and the combination of their results provide a full picture of the nanostructure induced by irradiation as each of them is sensitive to different aspects of the irradiation damage. The whole set of experimental conditions, both under neutron or ion irradiation, allows information to be obtained on the influence of the different variables involved, namely, Cr content, dose and irradiation temperature and, consequently, about fundamental processes occurring under irradiation. Furthermore, the results give the chance of comparing the effect of irradiation with two types of energetic particles, neutrons and ions, under similar conditions of dose and irradiation temperature. Therefore, three aspects will be addressed and discussed: (i) the description of the nanostructure produced by irradiation from the combination of results coming from different advanced characterisation techniques (ii) the influence of different variables and (iii) the comparison of the effect of irradiation with different energetic particles, i. e., neutrons and ions. Different features produced by irradiation have been revealed by every characterisation technique. In neutron irradiated materials, Cr-rich α' particles have been detected by SANS and APT in the supersaturated alloys Fe9Cr and Fe12Cr. Another family of Cr-

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rich nanoclusters containing also other impurity elements has been identified for each of the FeCr alloys. Solute enrichment to dislocation lines and grain boundaries has been revealed by APT. PAS has detected the damage in the form of vacancy clusters revealing that, when Cr is present, the vacancy clustering is suppressed. The interstitial clusters in the form of dislocation loops were observed and characterised by TEM. In the ion irradiated materials, dislocation loops by TEM and vacancy clusters by PAS have been detected. Regarding the solute redistribution, only one family of Cr-rich nanoclusters have been observed by APT. So far, no Cr-rich α' particles have been found in the conditions examined. Preliminary results from the EXAFS technique indicate that, at higher irradiation temperatures, the local Cr atomic environment is significantly changed, indicating a possible presence of short- range ordering.

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C-26

PREDICTION OF THE EFFECTS OF RADIATION FOR REACTOR PRESSURE VESSEL AND IN-VESSEL MATERIALS USING MULTI-SCALE MODELLING – 60 YEARS FORESEEN PLANT LIFETIME

A. Al Mazouzi1*, J. Sharples2 , M. Konstantinovic3, D. Moinereau1, D. Feron4

1EDF R&D, Ecuelles, Moret sur Loing Cedex, France

2SERCO assurance, Walton House, Warrington Cheshire, UK 3SCK.CEN, Mol, Belgium

4CEA, Saclay, France *Email: abderrahim.al-mazouzi@edf.fr

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and environmental effects, and in computer sciences encroach the development of multi-scale numerical tools able to simulate the material behavior in nuclear environment A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project [January 2002 – June 2008]. The FP7 Collaborative Project PERFORM 60 has been launched recently to pursue the further improvement of the existing tools, for reactor pressure vessel (RPV) steels and to initiate the development of similar multi-scale modelling tools to simulate the combined effects of irradiation and corrosion on the RPV internals. In fact, as the main goal of the project is to develop different mechanistic models at different levels of physics and engineering and to extend the state of knowledge in several scientific fields, the links between these different kinds of model are particularly difficult to deal with and need special techniques, which have yet to be fully realised. In this presentation, the models that have been developed with be presented and illustrated by examples on how they are used either to predict the behavior of the in-service components or to be used to assess the performance of the materials candidates for the next generation of nuclear power plants.

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C-27

PREDICTION OF IRRADIATION EMBRITTLEMENT OF VANADIUM ALLOYED LOW NICKEL STEEL FOR FUTURE REACTORS

A.Kryukov1*, L. Debarberis1, P. Hähner1, F. Gillemot2

1European Commission, Joint Research Centre - Institute for Energy and

Transport, Petten, The Netherlands 2Hungarian Academy of Science Atomic Energy Research Institute, Budapest,

Hungary *Email: al.kryukov2013@yandex.ru

Low alloyed steels have been developed and are extensively used as RPV materials for several decades. In particular, the second generation of WWER-440 materials (2CrMoV base metal and welds) with low content of impurities, have proven very high stability to radiation damage. This type of steel with low level of detrimental impurities presents acceptable radiation stability at least up to a neutron dose of 2 dpa at an irradiation temperature of 270 ºC. At irradiation temperatures above 400 ºC the embrittlement of it is very small (less than 30 °C DBTT shift observed at Tirr = 450 °C) mainly due to the dynamic recovery of radiation-induced defects. The comparison between pearlitic 2CrMoV and a ferritic-martensitic steel like EUROFER97 indicates that for neutron doses of 1.5 to 2 dpa and irradiation temperature of 300 ºC the DBTT shifts for WWER steel and EUROFER base material and welds are comparable. In the temperature range 350 to 500 ºC the radiation embrittlement level of both WWER and EUROFER steels are low. A thermal annealing of low temperature irradiated steel has been proposed as a promising method to predict results of high temperature irradiation embrittlement. Results indicate rather high temperature stability of WWER steel (base and weld metals) during the tempering up to 500 °C in unirradiated and irradiated conditions, making it an interesting option for high temperature applications.

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C-28 PNEUMATICALLY POWERED TESTING DEVICES FOR TESTING IN

REAL AND SIMULATED SERVICE ENVIRONMENTS

P. Moilanen1*, R. Novotny1, P. Hähner1 and S. Holmström2

1European Commission, Joint Research Centre - Institute for Energy and

Transport, Petten, The Netherlands 2VTT Technical Research Centre of Finland, Espoo, Finland

*Email: Kari-Pekka-Juhani.Moilanen@ec.europa.eu The requirements for materials testing in state-of-the-art and future energy production concepts have increased in complexity. A variety of well-defined and validated materials properties are required for design qualification. For instance damage rates related to corrosion, erosion, crack initiation and growth, fatigue, creep, irradiation and/or the combination thereof can be life limiting. In addition to classical mechanical testing such as tensile tests, advanced stress corrosion cracking, fatigue and creep tests, damage interaction and environmental compatibility tests are required. For instance the determination of the impact of in-pile irradiation on tensile and fatigue properties is of great interest but not readily available in the literature. To generate data for safe design in a variety of demanding environments it is necessary to enhance the capabilities of the testing systems. This paper aims to introduce the application areas of multifunctional high precision pneumatically powered testing systems. Different modifications of the basic pneumatic testing device have been utilized in a number of challenging environments, among others in laboratory simulated nuclear power plant water environments such as the BWR (Boiling Water Reactor), PWR (Pressurized Water Reactor) environments, SCW (Super Critical Water) and even in nuclear reactor in-pile testing. Furthermore, the conceptual design for a version suitable for testing in Liquid Lead environment up to 650 ºC is presented. The technological development path from single bellows tensile loading devices towards a more complex double bellows and double2bellows fatigue and combined tension/compression/internal pressure system is described.

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C-29

A NEW TEST SYSTEM OF AXIAL STRAIN CONTROLLED FATIGUE TEST WITH MINIATURE SPECIMEN

A. Nishimura1*, S.Nogami2 and K. Wakai3

1National Institute for Fusion Science, Oroshi, Toki, Japan

2Tohoku University, Tohoku, Japan 3Research and Development Center, Japan Atomic Energy Agency, Ibaraki,

Japan *Email: nishi-a@nifs.ac.jp

Fatigue property of a structural material for a fusion reactor is very important to ensure structural stability risks under dynamic operation and fast neutron irradiation. Since most of the structural materials will be irradiated by fast neutrons, generally speaking, the material becomes brittle with losing the elongation and the yield stress becomes higher. To investigate the fatigue property after neutron irradiation, a miniature specimen has been developed and used for fusion materials together with fission materials. An hour glass type specimen is also used for the study because the fracture position is able to be restricted. Hour glass type specimens do not provide the same fatigue life property as that shown by round bar type specimens. Also, the miniature specimen does not show the same fatigue property as the standard size specimen. To understand the reasons for these differences, a lot of attempts to develop the new test procedure for miniature size specimens but there are no excellent reports to describe the specimen size effect clearly. The key is considered to measure a longitudinal displacement of about 1 mm gage length keeping nano-scale precision. The displacement gages using strain gages were fabricated and tested. But it is not easy to produce the accuracy of nanometer range. In 2009, a new displacement equipment using laser was commercialized. It has an accuracy of 1 nm and the feed back is possible. Using the new laser displacement gages, a trial to assemble the new fatigue testing device for the miniature round bar specimens was performed. Although the improvement activities are still conducted, some fatigue test results are obtained. Before discussing the specimen size effect, the effect of machining surface, effect of number of grains on the cross section, effect of gage length and so on will be investigated in parallel. The paper will describe the development status of the new fatigue test system and present some fatigue data of JLF-1 which is a low activation ferritic/martensitic steel.

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C-30

STABILITY OF FERRITIC STEEL TO HIGHER DOSES; SURVEY OF RPV DATA AND COMPARISON WITH CANDIDATE MATERIALS FOR

FUTURE SYSTEMS

D.T. Blagoeva1*, L. Debarberis2, M. Jong1, P.ten Pierick1

1NRG, Westerduinweg 3, 1755 LE Petten, The Netherlands 2IET, JRC – EC, Westerduinweg 3, NL-1755 LE Petten, The Netherlands

*Email: blagoeva@nrg.eu Steels for future nuclear power reactor components as reactor pressure vessels (RPV), support structures and others will have to operate at higher temperatures than the current light water reactors – greater than 450oC and under the influence of higher neutron fluences, and in a different cooling environment. Therefore, special metallic materials, e.g. ferritic-martensitic steels are being developed for GEN IV reactor structures. On the other hand, the comprehensive experience of Light Water Reactors (LWR) operation also allows improved RPV materials to be offered as an option for many structures of the future reactors. Since decades low-alloyed steels have been developed and are extensively used as RPV materials for LWR. The irradiation data from LWRs during the past 30 years verify the radiation resistance of this class of steels. In particular, RPV materials with low content of impurities for the second and third generation of Nuclear Power Plants have proven very high resistance to radiation damage up to ~2 dpa and their response to radiation is very well understood as well as the effect of deleterious impurities. The newly developed advanced metallic structural materials for future fission and fusion reactors applications, e.g Eurofer97 (9%Cr) RAFM steel or ODS class steels, are demonstrating superior radiation damage resistance and a significantly higher corrosion resistance, improved physical and mechanical properties. However, manufacturing and joining of these steels in industrial scale is more costly and more specialized. Low alloyed RPV class steels could offer a similar performance at lower costs when used for vessel and other massive structural components. For neutron dose of 1.5-2 dpa and irradiation temperature 300oC different RPV steels and 9%Cr ferritic-martensitic steel show similar embrittlement behaviour (similar shifts in the transition temperature) when the level of impurities is low. However, the RPV steels have to be qualified for higher temperature applications and higher doses. The present paper compares several sets of surveillance and research data available for several RPV steels at doses up to 2 dpa and Eurofer97 data at doses up to 10 dpa to demonstrate the radiation stability of the first and their potential use as structural materials in the new generation nuclear systems.

83

C-31

EXPERIMENTAL CHARACTERIZATION OF THE MELTING OF REFRACTORY OXIDES: A LASER HEATING INVESTIGATION OF

NUCLEAR MATERIALS

L. Capriotti1,2, A. Quaini1,2, L. Luzzi2, R. Böhler1, K. Boboridis1, D. Manara1,3*

1Institute for Transuranium, Joint Research Centre of European Commission, Karlsruhe, Germany

2Department of Energy, "E.Fermi" Center for Nuclear Studies, Politecnico di Milano, Italy

*Email: dario.manara@ec.europa.eu The melting behaviour is a fundamental feature closely related to the structural and thermodynamic properties of a material, and constitutes therefore a crucial basic research subject. The melting point is also an important engineering parameter, as it largely contributes to the definition of the operational limits of a material element. However, experimental difficulties stemming both from the extreme temperature, implying high reactivity and fast kinetics, and vapour pressure conditions make the study of melting particularly challenging in refractory compounds. In many nuclear materials, the sample radioactivity enhances these difficulties even further. For this reason, experimental data in this domain are still rare and not always consistent. An overview of laser heating and fast pyrometry under "quasi containerless" conditions for high temperature material characterization is presented, with particular focus on melting studies. Some very recent results are shown, in particular on the so called "white oxides" (CaO, CeO2, ThO2) of interest for the nuclear industry. The high temperature study of these materials is particularly difficult and original because of their optical properties (transparency, reflection of laser light), and their poor chemical stability with respect to oxidation. The study was then completed with the successful investigation of the high temperature behaviour of mixed uranium-thorium dioxides. A critical comparison of experimental information obtained by traditional and advanced techniques is reported. Experimental data are described within a sounder theoretical background, essentially based on the optimization of phase diagrams.

84

C-32

ASSESSMENT OF VANADIUM ALLOYS FOR USE IN FISSION REACTORS

J.M. Gentzbittel1*,B. Riccetti1, M. V. Duquesnes2, M. Le Flem2

1CEA Grenoble, DRT/LITEN, 38054 GRENOBLE cedex 9-France

2CEA Saclay, DEN/DMN, 91191 Gif Sur Yvette cedex, France *Email: dario.manara@ec.europa.eu

According to technological challenges related to the development of fuel cladding for Gas Fast Reactors, GFR (SiC/SiC cladding) or Sodium Fast Reactors (ODS steels cladding), vanadium alloys turn out to be interesting candidates in the scope of future fission nuclear systems. Indeed, they are a good compromise between refractory features up to 700°C, compatibility with a fast spectrum, and reachable machining/shaping processes. Implementation of vanadium alloys in GFRs and SFRs is presently under consideration and must be estimated, especially regarding the new requirements of these applications (atmosphere, temperature, loading, etc.). In 2010, the manufacturing of a European CEA grade of vanadium alloy was launched: 30kg of V-4Cr-4Ti alloy were fabricated through GfE Metalle, Nuremberg, Germany. Based on fusion feedback, V-4Cr-4Ti reference grade was first chosen to validate the uneasy fabrication process linked to interstitial element sensitivity and potential pollution in master alloys. 7mm-plates were obtained after forging of the ingot, cold rolling with intermediate heat treatments and final machining. Final heat treatments were performed between 600°C and 1200°C to follow the recrystallization by hardness measurements and the grain size and Ti(O,C,N) precipitate observations. Together both microstructure characterization and the determination of tensile properties of recrystallized and stress relieved alloys have been performed up to 950°C. Creep tests were made up to 850°C under vacuum. The results and observations got at CEA will be presented; they are very consistent with previous investigations on V-4Cr-4Ti alloys by USA and Japan and highlight the good mechanical behaviour of this alloy at elevated temperature. These points out the very good quality of this European CEA grade, allowing to plan further optimisation and development.

85

C-33

DISLOCATION SINKS EFFICIENCY UNDER DIFFERENT TEMPERATURES AND APPLIED LOADS IN BCC IRON AND

VANADIUM

A.B. Sivak1*, V.M. Chernov2, V.A. Romanov3 and P.A. Sivak1

1National Research Centre “Kurchatov Institute”, Moscow, Russia 2A.A. Bochvar Institute of Inorganic Materials, Moscow, Russia

3A.I. Leipunsky Institute of Physics and Power Engineering, Obninsk, Russia *Email: sivak@rambler.ru

Simulation of self-point defects (SPDs) diffusion in elastic fields of edge and screw dislocations in <111>{110}, <111>{112}, <100>{100}, <100><011> slip systems under different applied loads at T = 293 – 1000 K was performed by object kinetic Monte Carlo method for bcc iron and vanadium crystals. Elastic interaction of SPDs considered as elastic dipoles (vacancy and self-interstitial atom in stable and saddle-point configurations) with dislocation and external stress fields was calculated by means of the anisotropic theory of elasticity. SPD jump directions were chosen randomly according to probabilities determined by corresponding energy barriers. Model crystallites containing a dislocation in its centre were right prisms of infinite length with square footing (footing side length, L, equals 200 – 400 a, where a is the lattice constant) with periodic boundary conditions applied to their side faces. Starting positions of SPDs were distributed randomly in the model crystallite. The trajectories were calculated until the SPD approached to the distance shorter than 3a from the dislocation and was regarded as absorbed. The dislocation sink efficiency, ξ, was calculated by known relation: ξ = 8L2/(a2<N>), where <N> is the average number of jumps performed by an SPD before its absorption. The number of simulated trajectories for each type of dislocation, external stress and SPD equaled 106 to obtain statistically reliable results. External stress tensors with nonzero off-diagonal components in crystallographic coordinate system with values in the range form –200 to +200 MPa were considered. Diagonal components do not affect the jump probabilities of SPDs therefore these components have no influence on the dislocation sink efficiency. Applied loads considerably change the dislocation sink efficiency. The effect strongly depends on the dislocation line direction and directions of principal axes of the external stress tensor. Obtained results allow one to evaluate the climbing rate of edge dislocations under irradiation. Considered features of SPDs behavior in dislocation and external fields can exert a significant influence on evolution of material microstructure under damage irradiation (swelling, creep, etc). __________________ The present work was funded by the Ministry of education and science of the Russian Federation (contracts #14.740.11.0162 and #02.740.11.0468).

86

C-34

MICROCHEMICAL EFFECTS IN IRRADIATED Fe-Cr ALLOYS AS REVEALED BY ATOMISTIC SIMULATION

D. Terentyev1, G. Bonny1, N. Castin1, L. Malerba1, E. Zhurkin2, M. Hou3,

K.Vörtler4, K. Nordlund4, E. Del Rio5

1Unit Structural Materials Modelling and Microstructure, Institute of Nuclear Materials Science, SCK•CEN, Mol, Belgium

2Experimental Nuclear Physics Department, Faculty of Physics and Mechanics, Saint-Petersburg State Polytechnical University, St. Petersburg, Russia

3Physique des Solides Irradiés et des Nanostrucutres CP234, Faculté des Sciences, Université Libre de Bruxelles, Bruxelles, Belgium

4Association EURATOM-Tekes, Department of Physics, University of Helsinki, Finland

5Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, Madrid, Spain *Email: dterenty@sckcen.be

Neutron irradiation produces defects in materials, the evolution of which changes their macroscopic properties. Defect production and evolution is expected to be influenced by the chemical composition of the material. In turn, the accumulation of defects in the material results in microchemical changes, which may induce further changes in macroscopic properties. In this work we review the results of recent atomic-level simulations conducted in Fe-Cr alloys, as model materials for high-Cr ferritic-martensitic steels, to address the following questions:

1. Is the primary damage produced in displacement cascades influenced by the Cr content? If so, how?

2. Does Cr change the stability of radiation-produced defects? 3. Is the diffusivity of cascade-produced defects changed by Cr content? 4. How do Cr atoms redistribute under irradiation inside the material under

the action of thermodynamic driving forces and radiation-defect fluxes? It is found that the presence of Cr does not influence the type of damage created by displacement cascades, as compared to pure Fe, while cascades do contribute to redistributing Cr, in the same direction as thermodynamic driving forces. The presence of Cr does change the stability of point-defects: the effect is weak in the case of vacancies, stronger in the case of self-interstitials. In the latter case, Cr increases the stability of self-interstitial clusters, especially those so small to be invisible to the electron microscope. Cr reduces also significantly the diffusivity of self-interstitials and their clusters, in a way that depends in a non-monotonic way on Cr content, as well as on cluster size and temperature; it has, however, no significant effect on single-vacancy diffusivity; moreover, the effect is negligible on the mobility of self-interstitial clusters large enough to become dislocation loops. Finally, Cr-rich precipitate formation is favoured in the tensile region of edge dislocations, while it appears not to be influenced by screw dislocations; prismatic dislocation loops (typically produced under irradiation) tend to be decorated by Cr. Cr has also tendency to accumulate at grain boundaries, while it tends to deplete in the proximity of free surfaces (at least in the absence of oxygen) and voids.

87

C-35 THE EFFECTS OF IRRADIATION DEFECT DISTRIBUTION AND THE STEEL COMPOSITIONS ON VOID DENUDED ZONE FORMATIONS

DURING NEUTRON IRRADIATION AND ELECTRON IRRADIATIONS

Y. Sekio1*, S. Yamashita1, N. Sakaguchi2 and H. Takahashi1,2

1 Research and Development Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki, Japan

2 Hokkaido University, Sapporo, Hokkaido, Japan *Email: sekio.yoshihiro@jaea.go.jp

The irradiation-induced void denuded zone (VDZ) formation behaviors near grain boundaries were studied in order to clarify the effects of irradiation defect distribution and chemical compositions of steels which were developed to retard their swelling because of the decrease in mobility of point defects during irradiation. The test materials were Fe-15Cr-15Ni model alloy and PNC316 stainless steel which was developed in order to improve the void swelling resistance. These steels were neutron-irradiated in the experimental fast reactor JOYO at temperature range from 749 K to 775 K with a range of fast neutron fluence 18-103 dpa and electron-irradiated separately using the 1MeV high voltage electron microscopy (HVEM) at temperature range from 723 K to 773 K with the electron dose up to 3.6 dpa. The VDZ widths of each steel were evaluated by the TEM microstructural observations after irradiation. The experimental results showed that there was no noticeable difference in the VDZ widths of the Fe-15Cr-15Ni steels in spite of discrepancy in irradiation defect distribution. This suggests that the VDZ width would not be basically influenced by the difference of irradiation conditions, such as irradiation defect distribution and dose rates. However, the VDZ width of the PNC316 steel was narrower than that of the Fe-15Cr-15Ni steel after neutron irradiation. This result indicates that the behavior of point defects accumulation near grain boundaries during irradiation should be different between Fe-15Cr-15Ni and PNC316 steels. Therefore, it is considered that VDZ formation would be affected by chemical compositions and that the decrease of VDZ width would be related to increase of the void swelling resistance.

88

C-36

USE OF SELF-ION IRRADIATION TO STUDY VOID SWELLING AND PHASE STABILITY IN ADVANCED FERRITIC-MARTENSITIC STEELS

V. N. Voyevodin1, V.V. Bryk1, O.V. Borodin1, V.V. Melnichenko1, F.A.

Garner2

1Kharkov Institute of Physics and Technology, Kharkov, Ukraine 2Radiation Effects Consulting, Richland WA, USA

*Email: voyev@kipt.kharkov.ua Ferritic-martensitic alloys and oxide-dispersion-strengthened variants of these alloys are strong favorites to replace swelling-prone austenitic alloys to reach very high neutron exposures. However, it is currently impossible to obtain more than 20 dpa per year in existing fast reactors for irradiation of commercial or developmental structural alloys. Currently there is strong demand for multiple irradiations, especially on developmental alloys, to doses greater than 200 dpa, perhaps even to 600 dpa, all results to be delivered within months, rather than over decades. For some types of radiation-induced property changes charged particle irradiation at accelerated damage rates can be confidently employed to examine the consequences of radiation on this alloy class. Foremost of these property changes are void swelling, dislocation evolution and phase stability. Ion irradiation, if conducted and evaluated properly can be used to determine the nature of the swelling law (nonlinear, linear or bilinear) and the maximum post-transient or steady-state swelling rate. Our studies using 1.8 MeV Cr+ ion irradiation at 1x10-2 dpa/sec on a variety of ferritic alloys show that this alloy class exhibits a bilinear behavior with a post-transient swelling rate of ~0.2%/dpa consistent with earlier predictions based on theory and neutron irradiation data. Of particular importance has been our finding that ferrite grains swell rather easily but decomposed martensite grains with high densities of various carbide phases resist the onset of swelling to much higher doses. While the experimentally-derived temperature dependence of swelling and the duration of the transient regime cannot be confidently determined using ion irradiation, it is possible to examine the microstructural and microchemical evolution that precedes and accompanies void swelling, especially in the transition regime between the pre-transient and post-transient swelling regimes. A review of our results in the range of 100-600 dpa is presented for a variety of alloys (decomposed martensite, duplex ferrite/decomposed martensite, ferrite with dispersed oxides). Also addressed are our findings concerning the optimum irradiation and examination conditions needed to produce reliable results, focusing on injected interstitials, surface effects, rastered vs. defocused beams, etc.

89

C-37

SYNERGISTIC EFFECT OF HE AND DISPLACEMENT CASCADE IN FeCr ALLOYS STUDIED AT ATOMISTIC SCALE

X. He*, J. Yu, P. Yang, W. Yang

China Institute of Atomic Energy, Beijing, China

*Email: hexinfu@ciae.ac.cn Reduced activation ferritic/martensitic (RAFM) steels are considered as structural materials for fusion power systems. 14 Mev neutron irradiation will result in the production of substantial amount of He (up to ~2000 appm at end-of-life dose i.e. ~200 dpa). Being essentially insoluble and highly mobile as interstitial, He resides at lattice defects such as vacancy clusters, voids, dislocations, grain boundaries, lath boundaries and particle–matrix interfaces, where an extra volume is available. The synergistic effects of displacement damage and helium on mechanical properties and microstructure in FeCr model alloys and RAFM steels, such as F82H, 9Cr-2WVTa are one of the special problems in the design of fusion materials. This is why a significant experimental and theoretical effort is devoted to study properties of He in RAFM steels, Fe-based binary alloys and pure Fe. Systematic Molecular dynamics (MD) simulation of displacement cascades in Fe–0.1%He and Fe-10Cr-0.1%He are used to investigate the synergistic effects of displacement cascades and He on the irradiation damage production. The simulations were performed in the temperature range 10 ~ 873 K, applying a recently proposed set of interatomic potentials for Fe-Cr-He system. The effects of substitutional / interstitial He on the defect productivity, defects configuration and cluster formation were characterized in detail. The effects of Cr concentration and irradiation temperature on the irradiation induced defects were studied also.

90

C-38

EFFECT OF RADIATION-INDUCED DAMAGE ON DEUTERIUM RETENTION IN TUNGSTEN AND TUNGSTEN COATINGS

O. V. Ogorodnikova* and K. Sugiyama

Max-Planck-Institut für Plasmaphysik, EURATOM Association; Boltzmannstr. 2,

D-85748 Garching, Germany *Email: olga.ogorodnikova@ipp.mpg.de

Tungsten, its alloys and W coatings are primary candidates for plasma-facing materials (PFMs) in fusion reactors. The plasma-facing components of fusion reactors will be exposed to high flux of deuterium (D) and tritium (T) under the irradiation by 14 MeV neutrons (n) together with high rate of helium (He) and hydrogen (H) production due to transmutation reactions. Radiation-induced defects produced by n-irradiation act as trapping sites for hydrogen isotopes, and, consequently, T inventory in n-irradiated W materials is an important issue due to the radioactivity of tritium and for the plasma fuel balance. Experimental data on the influence of displacement damage on hydrogen accumulation, recycling and permeation in materials are very scarce, and the detailed theoretical understand of hydrogen transport through damaged materials is incomplete. These data are strongly required not only for ITER, but also for future fusion power reactors such as DEMO. To understand and predict T transport and retention in complicate irradiation conditions, interaction of low energy deuterium with polycrystalline tungsten (W) and nanostructured W coating produced by Combined Magnetron Sputtering and Ion Implantation (CMSII) technology is investigated under well-defined laboratory conditions. In this paper, we highlighted a difference in the deuterium retention in a bulk W and in dense nanostructured W coatings as well as a difference in radiation-induced damage produced in a bulk W and in W coatings. To simulate the fast neutron damage produced in fusion reactors, W was irradiation by 20 MeV W ions. Following to the damage production by the irradiation at different dpa levels, samples were exposed to low-energy deuterium plasma in IPP. The rate of deuterium decoration of radiation-induced damage depends on temperature, ion energy and ion flux. The effect of temperature was investigated by variation from room temperature to 800 K and ion energy from 5 to 200 eV. The D retention in each sample was subsequently analyzed by various methods such as nuclear reaction analysis (NRA) for the depth profiling up to 6 μm and thermal desorption spectroscopy for the determination of total amount of retained D. The diffusion model with dynamic trap formation during irradiation was applied to the assessment of binding energies of deuterium with natural and radiation-induced defects and density of the defects. The nature of traps produced by heavy ion implantation in W, and its influence on retention behaviour under fusion reactor conditions is discussed.

91

C-39

STUDY ON HEAVY IRRADIATION DAMAGE IN MATERIALS FOR POSSIBLE NUCLEAR FUSION APPLICATION IN HIGH NEUTRON

FLUX FAST REACTOR OF JOYO

M. Watanabe1, T. Shikama1*, T. Asaga2, S. Yamashita2, M.Itoh2

1Institute for Materials Research, Tohoku University, 980-8577 Sendai, Japan 2Japan Atomic Energy Agency, Oarai, 311-1394 Oarai, Japan

*Email: shikama@imr.tohoku.ac.jp Heavy irradiation effects upon property changes in nuclear fusion materials must be assessed adequately for the development of nuclear fusion reactors. There, more than a few 10 dpa irradiation is needed by neutrons, desirably in nuclear fusion relevant irradiation conditions. A high flux fast fission reactor will be an only irradiation bed for realizing the conditions except for the nuclear transmutation aspects by high energy neutrons of more than 1 MeV, assuming that the irradiation conditions are well controlled and the obtained results could be analyzed reliably in comparison with other irradiation results obtained in more well-defined conditions. JOYO, the experimental fast reactor, whose fast (E>0.1MeV) neutron flux is higher than 10x1019 n/m2s, which is mandatory for the heavy irradiation damage study for the nuclear fusion materials, has developed well-controlled irradiation techniques such as MARICO-II, as well as the irradiation rig, called Shuttle Rig convenient for frequent iterations between reactor irradiations and post irradiation examinations (PIEs). Also, the JAEA-Oarai, Tohoku University, and the NFD are collaborating closely to establish a hot-laboratory network in Oarai area, for variety of PIEs needed for advanced materials studies, such as, mechanical tests with from the standard size to miniature size specimens, a variety of micro to nano analysis techniques. The paper will describe general features of the ability of studies of heavy irradiation effects in JOYO and its relating hot laboratories, and it will report some experimental results on ceramic and metallic materials for possible nuclear fusion applications. For the case of metallic materials, correlation of heavy radiation effects among different irradiation beds, such as charged particle irradiations, neutrons, and computer simulations, is being well established and now the efforts are toward the comprehensive understandings of the radiation effects including those of nuclear fusion relevant conditions. However, the obtained results on ceramic materials show that the behaviors of heavily irradiated materials in JOYO are definitely different from those irradiated by the charged particles. Also, the results are implying that the general model of the microstructural evolutions in metallic material systems may not be applied to the ceramic materials, where the strong electrical interactions among radiation induced defects will be strong and cannot be compatible with the assumed stability of radiation induced defects in the metallic materials.

92

C-40

RADIATION STABILITY OF THE ODS ALLOYS AGAINST SWIFT HEAVY ION IMPACT

V.A. Skuratov1*, V.V. Uglov2, J. O'Connell3, A.S.Sohatsky1, J.H. Neethling3,

S.V.Rogozhkin4

1FLNR, JINR, Dubna, Russia 2Belarusian State University, Minsk, Belarus

3CHRTEM, NMMU, Port Elizabeth, South Africa 4ITEP, Moscow, Russia *Email: skuratov@jinr.ru

Recent TEM examinations of the ODS alloys irradiated with high energy Xe and Kr ions have revealed strong sensitivity of oxide particles to dense ionization [1,2]. Experimental data about of such effects are of considerable practical value in view of simulation of fission product impact. In this report we present the results of SEM/EDX and TEM studies of Cr16(VNIINM), KP-4 and KP-1,2,3 ODS alloys irradiated with 167 MeV Xe and 700 MeV Bi ions. First, the changes in surface topography of Cr16 steel induced by Xe ions with fluences ranged from 2.78×1014 to 2.56×1015 cm-2 are discussed. It was found that inelastic sputtering leads to decomposition of Y2O3 particles accompanied by formation of Y-rich phase. Also, the layering of Cr(Fe,W) particles with formation of Fe- and Cr-rich regions is observed. These results indicate that surface effects of dense ionization should be taken into account for simulation of fission fragment impact. Second part of report is devoted to oxide particle stability against high energy Xe and Bi ion irradiation. The TEM analysis have shown that significant number of oxide nanoparticles remain crystalline even at highest Xe and Bi ion fluences – 1.5×1015 cm-2 and 1.5×1013 cm-2, correspondingly, when ion track regions are multiply overlapped. The targets used in our irradiation experiments were the bulk materials in comparison with pre-thinned specimens used in works [1,2]. The lower sensitivity of oxides particles to dense ionization found in our study may relate the peculiarities of the energy relaxation processes under swift heavy ion bombardment in bulk and thin (tens nanometers thickness) materials. References. [1] J. Ribis et al., J. Nucl. Mater. 417 (2011) 262. [2] I. Monnet, C. Grygiel, M. L. Lescoat, J.Ribis Amorphization of oxides in

ODS steels/materials by electronic stopping power, J. Nucl. Mater. in press.

93

C-41 RADIATION TOLERANCE OF NANOSTRUCTURED ZRN COATINGS

AGAINST SWIFT HEAVY ION IRRADIATION

A. Janse van Vuuren1*, V. A. Skuratov2, V.V. Uglov3, J.H. Neethling1, S.V.Zlotski3

1CHRTEM, NMMU, Port Elizabeth, South Africa

2FLNR, JINR, Dubna, Russia 3Belarusian State University, Minsk, Belarus

*Email: arnojvv@gmail.com Zirconium Nitride is one of the ceramics under consideration as a candidate inert matrix fuel host for fast reactors or accelerator-driven sub-critical systems [1, 2]. It has been demonstrated that ZrN has a high tolerance to, low-energy, heavy and light ion bombardment [3, 4]. At the same time, limited data is available regarding the radiation resistance of ZrN to fission fragment bombardment. Information concerning the radiation defects in nano-structured ZrN, irradiated with swift heavy ions, is also limited. The aim of this study is the structural analysis of nanocrystalline ZrN coatings irradiated with high energy Xe and Bi ions, in order to simulate radiation damage resulting from fission fragment bombardment. ZrN layers, with nanocrystalline microstructure (average crystallite size of ~ 4 nm), of different thicknesses (0.1, 3, 10 and 20 μm) were produced via vacuum arc-vapour deposition. These samples were irradiated with 167 MeV Xe and 695 MeV Bi ions to fluences in the range from 3×1012 to 2.6×1015 cm-2 for Xe and 1012 to 1013 cm-2 for Bi. The irradiated samples were subsequently studied by XRD and TEM techniques. The XRD analysis revealed no apparent changes in the phase composition after irradiation. This finding was true even for particles with the highest specific ionization energy loss of 49 keV/nm for monatomic particles in ZrN. This result is indicative of the good structural stability of nanocrystalline layers under dense ionization effects. The XRD results are in agreement with TEM studies, which show no crystalline to amorphous phase transition as well as the absence of any specific radiation damage effects, which may be ascribed to latent track formation. The peculiarities of stress accumulation in ZrN layers of different thicknesses irradiated with swift heavy ions for a range of fluences will also be discussed. References. [1] H. Kleykamp, J. Nucl. Mater. 275 (1999) 1. [2] M. Streit, F. Ingold, M. Pouchon, L.J. Gauckler, J.-P. Ottaviani, J. Nucl.

Mater. 319 (2003) 51. [3] G.W. Egeland, J.A Valdez, J.G. Swadener, B. Oliver, K.J. McClellan, S.A.

Maloy, K.E. Sickafus, G. Bond, Heavy Ion Irradiation Effects in Zirconium Nitride, in: Proceedings of the 2004 International Congress on Advances in Nuclear Power Plants, ICAPP’04, vol. 4225, 2004, 2023.

[4] Yong Yang, Clayton A. Dickerson, Todd R. Allen, J. Nucl. Mater. 392 (2009) 200.

94

C-42 INVESTIGATIONS ON THE JOINING OF 9-20CR ODS AND NON-ODS STEELS APPLYING DIFFUSION, ELECTRON BEAM AND FRICTION

STIR WELDING

L. Commin1, U. Jäntsch1, M. Klimenkov1, R. Lindau1*, A. Möslang1, P. Norajitra1, L. Bergmann2, J.F. dos Santos2

1Karlsruhe Institute of Technology, Institute for Applied Materials, Karlsruhe,

Germany 2Helmholz-Zentrum Geesthacht, Institute of Materials Research, Geesthacht,

Germany *Email: rainer.lindau@kit.edu

Reduced activation ferritic martensitic (RAFM) oxide dispersion strengthened (ODS) steels like the 9Cr Eurofer-ODS and higher chromium containing ODS steels are candidate materials to be used as structural materials in blanket as well as divertor applications of advanced future power nuclear fusion reactors. Their use would allow increasing the operational temperature to 650-750°C compared to standard Eurofer, the European reference steel for DEMO structures. 9Cr ODS as well as ferritic ODS steels with higher Cr contents like PM2000 are also of interest for advanced fission reactor application. One drawback of ODS steels is difficult joining. Standard fusion welding techniques like electron beam welding (EBW) can only be applied in regions of lower demands since the melting process leads to a loss of strength due to coarsening and agglomeration of the strengthening nanometric ODS particles. The applicability of diffusion welding (DW) for ODS plating of components made of Eurofer steel was successfully demonstrated. Friction stir welding (FSW), as another solid-state joining process, could be an alternative way to join ODS alloys while preserving the advantageous microstructure. Similar and dissimilar joints of Eurofer-ODS and Eurofer have been fabricated applying DW, EB and FSW techniques. A 20Cr ODS steel (PM2000) was friction stir welded to assess the behaviour of purely ferritic oxidation resistant ODS steels which could be interesting also for GenIV fission reactors. Specific post-weld heat treatments (PWHT) were applied to optimise the impact and tensile properties which were measured utilizing small scale specimen test technology. The resulting microstructure was characterized by means of optical microscopy, scanning electron microscopy (SEM) and ion-beam-imaging in a Dual-Beam-SEM focused ion beam (FIB). Low magnification and analytical transmission electron microscopy was performed using a 200 kV Tecnai 20 FEG TEM with scanning unit (STEM) and high-angle-annular-dark field (HAADF) detector and EDX. The mechanical behaviour can be well correlated with the micro- and nanostructural changes introduced by the different welding processes and post-weld heat treatments.

95

List of participants

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ABDERRAHIM AL MAZOUZI EDF R&D Avenue des Renardières-Ecuelles 77818 MORET SUR LOING (France) tel: +33160736219 E-mail: abderrahim.almazouzi@edf.fr STEFFEN ANTUSCH KIT H.-v.-Helmholtz-Platz 1 76344 EGGENSTEIN (Germany) E-mail: Steffen.antusch@kit.edu FRANK BERGNER HZDR Bautzner Landstrasse 400 01328 DRESDEN (Germany) E-mail: f.bergner@hzdr.de DARINA BLAGOEVA NRG Westerduinweg 3 1755 ZG PETTEN (Netherlands) tel: 0031224568341 E-mail: blagoeva@nrg.eu NATALIA BRIKOTNINA Radiation Effects Consulting 2003 Howell Avenue 99354 RICHLAND (United States of America) tel: 1-509-521-1633 fax: 1-509-946-5542 E-mail: nataliabrikotnina@att.blackberry.net LUCA CAPRIOTTI Politecnico di Milano via vare 7/9 20158 MILANO (Italy) tel: 3497139553 E-mail: luca.capriotti@mail.polimi.it SUBASH CHANDER CHETAL Indira Gandhi Centre for Atomic Research DAE Campus 603102 KALPAKKAM (India) tel: +91 044 27480240 fax: +91 044 27480060 E-mail: dir@igcar.gov.in TAKUMI CHIKADA The University of Tokyo 2-11-16 Yayoi 113-8656 BUNKYO-KU (Japan) tel: +81-3-5841-7420 fax: +81-3-5841-7420 E-mail: chikada@nuclear.jp LORELEI COMMIN KIT H von Helmoltz platz 76133 EGGENSTEIN LEOPOLDSHAFEN (Germany) E-mail: lorelei.commin@kit.edu

YANN DE CARLAN CEA CEA Saclay 91191 GIF SUR YVETTE (France) tel: 33169086175 fax: 33169087130 E-mail: yann.decarlan@cea.fr MONICA FERRARIS politecnico di torino corso duca degli abruzzi 24 10129 TORINO (Italy) tel: +39 011 090 46 87 E-mail: monica.ferraris@polito.it MICHAEL FLUSS LLNL East Avenue 94550 LIVERMORE, CA (United States of America) tel: +1 925-423-6665 fax: none E-mail: fluss1@llnl.gov PIERRE-ERIC FRAYSSINES CEA 17 rue des Martyrs 38054 GRENOBLE CEDEX 9 (France) tel: 0438782859 E-mail: pierre-eric.frayssines@cea.fr FRANCIS GARNER Radiation Effects Consulting 2003 Howell Avenue 99354 RICHLAND (United States of America) tel: 1-509-521-1633 E-mail: frank.garner@dslextreme.com SERGUEI GAVRILOV SCK·CEN Boeretang 200 2400 MOL (Belgium) tel: +32 14 33 30 67 fax: +32 14 32 12 16 E-mail: sgavrilo@sckcen.be MARK GILBERT CCFE Culham Science Centre OX14 3DB ABINGDON (United Kingdom) E-mail: mark.gilbert@ccfe.ac.uk FERENC GILLEMOT MTA EK Konkoly Thege 29-33 1125 BUDAPEST (Hungary) tel: +3613922222 ext 1420 fax: +3623523690 E-mail: gillemot.ferenc1@t-online.hu

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SEHILA GONZALEZ DE VICENTE EFDA Boltzmannstrasse 85748 GARCHING BEI MUNCHEN (Germany) tel: 00498932994209 fax: 00498932994312 E-mail: sehila.gonzalez@efda.org XINFU HE China Institute of Atomic Energy P.O. box: 275-51,Xinzhen, Fangshan, Beijing 102413 BEIJING (China) tel: +861069357161 fax: +861069357161 E-mail: hexinfu@ciae.ac.cn MERCEDES HERNANDEZ-MAYORAL CIEMAT Avenida Complutense, 40 28040 MADRID (Spain) tel: +34913466618 fax: +34913466661 E-mail: m.mayoral@ciemat.es DAVID HOELZER Oak Ridge National Laboratory 1 Bethel Valley Road; MS6136 37831-6136 OAK RIDGE (United States of America) tel: 1-865-574-5096 fax: 1-865-241-3650 E-mail: hoelzerd@ornl.gov JAN HOFFMANN Karlsruhe Institute of Technology Herrmann-von-Helmholtz-Platz 1 76744 EGGENSTEIN-LEOPOLDSHAFEN (Germany) E-mail: j.hoffmann@kit.edu LUKE HSIUNG Lawrence Livermore National Laboratory 7000 East Avenue 94551 LIVERMORE (United States of America) tel: +1 925-4243125 fax: +1 925-4243815 E-mail: hsiung1@llnl.gov PETER HÄHNER Joint Research Centre Westerduinweg 3 17755 LE PETTEN (Netherlands) tel: 0031224565217 fax: 0031224565627 E-mail: peter.haehner@ec.europa.eu VICTOR IGNATIEV NRC Kurchatov Institute Kurchatov sq. 1 123182 MOSCOW (Russian Federation) tel: +74991967130 fax: +74991966172 E-mail: ignatiev@vver.kiae.ru

JINSUNG JANG Korea Atomic Energy Research Institute 989-111 Daedeok-daero, Yuseong 305-353 DAEJEON (South Korea) tel: 82-42-868-2376 fax: 82-42-868-8549 E-mail: jjang@kaeri.re.kr ARNO JANSE VAN VUUREN Nelson Mandela Metropolitan University CHRTEM Building, NMMU South Campus, University Way 6001 PORT ELIZABETH (South Africa) tel: +27415044366 E-mail: arnojvv@gmail.com TAMMANA JAYAKUMAR Indira Gandhi Centre for Atomic Research Metallurgy and Materials Group 603102 KALPAKKAM (India) tel: 00914427480107 fax: 00914427480075 E-mail: tjk@igcar.gov.in GENTZBITTEL JEAN MARIE CEA Grenoble 17,avenue des Martyrs 38054 GRENOBLE (France) tel: 33 438789501 fax: 33 438785891 E-mail: jean-marie.gentzbittel@cea.fr JEOUNG HAN KIM Korea Institute of Materials Science Changwondaero 797 642-831 CHANGWON (South Korea) tel: +82-55-280-3372 fax: +82-55-280-3255 E-mail: kjh1754@kims.re.kr TAE KYU KIM KAERI Daedeok-daero 989-111 305-353 DAEJEON (South Korea) E-mail: tkkim2@kaeri.re.kr RADIM KOPRIVA Nuclear Research Institute Rez Husinec-Rez 130 25068 REZ (Czech rep.) E-mail: kor@ujv.cz ALEXANDER KRYUKOV Joint Research Centre Westerduinweg 3 1755 ZG 1815 CK (Netherlands) E-mail: al.kryukov2013@yandex.ru MILOS KYTKA Nuclear Research Institute Rez Husinec-Rez 130 25068 REZ (Czech rep.) E-mail: kyt@ujv.cz

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CHUNJING LI Institute of Plasma Physics,CAS Shushan Lake Road 350# 230031 HEFEI (China) tel: +865515592424 fax: +865515591397 E-mail: lcj@ipp.ac.cn RAINER LINDAU Karlsruhe Institute of Technology, Institute for Applied Materials, KARLSRUHE (Germany) rainer.lindau@kit.edu NATALIA LUZGINOVA Nuclear Research and consultancy Group Westerduinweg 3 1755ZG PETTEN (Netherlands) E-mail: luzginova@nrg.eu LORENZO MALERBA SCK-CEN Boeretang 200 2400 MOL (Belgium) E-mail: lmalerba@sckcen.be DARIO MANARA European Commission JRC Institute for Transuranium Elements Materials Research Unit P.O. Box 2340 D-76125 Karlsruhe, Germany +49 (0) 7247 951 129 E-mail: dario.manara@ec.europa.eu PEKKA MOILANEN Joint Research Centre Westerduinweg 3 1755 ZG PETTEN (Netherlands) E-mail: kari-pekka-juhani.moilanen@ec.europa.eu WILLY MUNTJEWERF Joint Research Centre Westerduinweg 3 1755 ZG PETTEN (Netherlands) tel: 0031224565109 fax: 0031224565627 E-mail: willy.muntjewerf@ec.europa.eu HORVÁTH MÁRTA MTA Centre for Energy Research Konkoly Thege M. ut 29-33 1121 BUDAPEST (Hungary) E-mail: horvath.marta@energia.mta.hu ARATA NISHIMURA National Institute for Fusion Science 322-6 Oroshi 509-5292 TOKI (Japan) tel: +81-572-58-2118 fax: +81-572-58-2676 E-mail: nishi-a@nifs.ac.jp

RADEK NOVOTNY Joint Research Centre Westerduinveg 3 1755 LE PETTEN (Netherlands) E-mail: Radek.Novotny@ec.europa.eu OLGA OGORODNIKOVA Max-Planck Institute fuer Plasmaphysik Boltzmannstr.2 85748 GARCHING (Germany) tel: 00498932991919 fax: 00498932992279 E-mail: olga.ogorodnikova@ipp.mpg.de JIN-JU PARK Korea Atomic Energy Research Institute (KAERI) Dukjin-dong, Yuseong-gu 305-353 DAEJEON (South Korea) E-mail: jinjupark@kaeri.re.kr ANDREA QUAINI Politecnico di Milano Piazza Leonardo da Vinci 20133 MILAN (Italy) E-mail: quaio87@gmail.com DANIEL RILEY Institute of Materials Engineering Australian Nuclear Science & Technology Organisation Locked Bag 2001 Kirrawee DC NSW 2232 Tel: + 02 9717 9337 E-mail: Daniel.Riley@ansto.gov.au ALEXANDER RYAZANOV National Research Centre "Kurchatov Institute" Kurchatov Sq.1 123182 MOSCOW (Russian Federation) tel: +74991969177 fax: +74954214598 E-mail: ryazanov@cern.ch YOSHIHIRO SEKIO Japan Atomic Energy Agency 4002, Narita-cho, Oarai-machi 311-1393 IBARAKI (Japan) tel: +81-29-267-4141 fax: +81-29-266-3713 E-mail: sekio.yoshihiro@jaea.go.jp MARTA SERRANO CIEMAT Avda de ka Complutense 40 28040 MADRID (Spain) tel: +34913466030 fax: +34913466661 E-mail: marta.serrano@ciemat.es TATSUO SHIKAMA Tohoku University 2-1-1 Katahira, Aobaku 980-8577 SENDAI (Japan) tel: 81222152060 fax: 81222152061 E-mail: shikama@imr.tohoku.ac.jp

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ALEXANDER SIVAK NRC "Kurchatov institute" 1, Kurchatov sq. 123182 MOSCOW (Russian Federation) E-mail: sivak@rambler.ru VLADIMIR SKURATOV Joint Institute for Nuclear Research Joliot-Curie 6 141980 DUBNA (Russian Federation) tel: +7 (49621) 65-059 fax: +7 (49621) 65-146 E-mail: skuratov@jinr.ru VLADIMIR SLUGEN Slovak University of Technology Ilkovicova 3 81219 BRATISLAVA (Slovakia) E-mail: vladimir.slugen@stuba.sk MIKHAIL SOKOLOV ORNL P.O.Box 2008 37831 OAK RIDGE (United States of America) tel: 1-865-574-4842 fax: 1-865-574-6095 E-mail: sokolovm@ornl.gov HEISHICHIRO TAKAHAHSI Hokkaido university kita-ku,kita-13,nishi-8 +81-060-86 SAPPORO (Japan) tel: +81-11-706-6767 fax: +81-757-35-37 E-mail: takahash@ufml.caret.hokudai.ac.jp A-A. FARHAD TAVASSOLI CEA DMN/Dir 91191 GIF-SUR-YVETTE (France) tel: 0169086021 fax: 0169088070 E-mail: farhadt@orange.fr

LOUISE TOUALBI CEA CEA Saclay 91191 GIF-SUR-YVETTE (France) E-mail: louise.toualbi@gmail.com VALENTYN TSISAR Physical-Mechanical Instsitute of NASU 5, Naukova St. 79601 LVIV (Ukraine) E-mail: valentyn_tsisar@ukr.net SHINICHIRO YAMASHITA Japan Atomic Energy Agency 4002, Narita-cho, Oarai-machi 311-1393 IBARAKI (Japan) tel: +81-29-267-1919 fax: +81-29-266-3713 E-mail: yamashita.shinichiro@jaea.go.jp OLHA YELISEYEVA Physical-Mechanical Institute of NASU 5, Naukova St. 79601 LVIV (Ukraine) E-mail: olgayeliseyeva@ipm.lviv.ua ANDREJ ZEMAN IAEA Wagramerstr.5 1400 VIENNA (Austria) E-mail: a.zeman@iaea.org STEVEN ZINKLE Oak Ridge National Laboratory 1 Bethel Valley Rd 37831 OAK RIDGE, TENNESSEE (United States of America) tel: (865) 576-5785 fax: (865) 574-8872 E-mail: zinklesj@ornl.ogv

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European Commission

EUR 25326 – Joint Research Centre – Institute for Energy and Transport

Title: Development of new Structural Materials for Advanced Fission and Fusion Reactor Systems - Book of abstracts of the 2nd

joint IAEA-EC topical meeting

Editors: A. Zeman, P. Haehner

Luxembourg: Publications Office of the European Union

2012 – 101 pp. – 21.0 x 29.7 cm

EUR --- Scientific and Technical Research series --- ISSN 1831-9424 (online) - ISSN 1018-5593 (print)

ISBN 978-92-79-24908-2 (pdf)

ISBN 978-92-79-24907-5 (print)

doi:10.2790/52879

Abstract

Further to a successful Topical Meeting on "Development of New Structural Materials for Advanced Fission and Fusion Reactor

Systems" jointly organised by IAEA and JRC-Institute for Energy and Transport and held in Oct. 2009 at the premises of Fusion

for Energy in Barcelona, the 2nd Joint IAEA-EC Topical Meeting on the same subject took place on 16 – 20 April 2012 at the JRC

in Ispra, Italy. The Topical Meeting has again provided a well received platform for detailed presentations, technical discussions

and exchange of results in the specific areas of relevance to materials performance assessment and qualification for advanced

fission on the one hand, and thermo-nuclear fusion systems on the other hand. In fact, the Topical Meeting has achieved its

objective to gather experts from both scientific communities, in order to develop synergies between the fields of research.

Following keen demand for participation, the Topical Meeting was limited to 65 international delegates from 20 countries, with

the strongest participation from the USA, Russia, Japan, Germany, France, and The Netherlands. The present report contains the

collection of abstracts of the papers presented, while full size papers will be published as a Themed Issue of Journal of Nuclear

Materials.

z

As the Commission’s in-house science service, the Joint Research Centre’s mission is to provide EU policieswith independent, evidence-based scientific and technical support throughout the whole policy cycle. Working in close cooperation with policy Directorates-General, the JRC addresses key societal challenges while stimulating innovation through developing new standards, methods and tools, and sharing andtransferring its know-how to the Member States and international community. Key policy areas include: environment and climate change; energy and transport; agriculture and foodsecurity; health and consumer protection; information society and digital agenda; safety and security including nuclear; all supported through a cross-cutting and multi-disciplinary approach.

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