Domaine concerné, contexte et principaux enjeux

Centre of Excellence for Nuclear Materials

Workshop

Materials Innovation for Nuclear

Optimized Systems

December 5-7, 2012

CEA – INSTN Saclay, France

Workshop Materials Innovation for Nuclear Optimized Systems December 5-7, 2012, CEA – INSTN Saclay, France

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Content

Scope and objective of the workshop

Workshop program

List of the speakers and lists of local scientific committee members, organizers, sponsors

Opening address

Abstracts of keynote and lecture speakers (sorted by order of sessions)


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Scope and objective of the workshop Complex systems for nuclear-based electricity may be exposed to severe service conditions (irradiation, temperature, mechanical loading, corrosive environments). It is thus essential to rely on high-performance and durable materials and structures. In order to extend the service life of reactors and to improve NPP availability, research must be performed to guarantee that they can be done safely. Optimizing current and future claddings and fuels, and designing new materials for 4th generation reactors call for innovative breakthroughs to increase the reliability of predictive models concerning nominal conditions and potential accident scenarios. The MINOS (Materials Innovation for Nuclear Optimized Systems) Centre of excellence for nuclear materials was launched in 2011 by the Nuclear Energy Division (DEN) of CEA. MINOS consists of 600 scientists involved in basic and applied research on nuclear materials in areas such as chemistry, physics, mechanics and behavior under irradiation. It covers a broad spectrum of R&D activities in the field of materials science dedicated to nuclear applications (nuclear reactors, waste management) and involves the other CEA main divisions. As an international reference, MINOS contributes to the consolidation of both scientific and technological potential of CEA. MINOS sustains strategic research partnerships and innovative research programs related to material (elaboration, characterization, design and modeling/simulation) and structure behavior in severe environments. The center is committed to collaborative partnerships to promote and develop training and educational activities in materials science. This initiative has allowed CEA to reaffirm its position as a world leader in nuclear materials R&D and to demonstrate a strong innovative capacity and ability to execute and supervise research programs for industrial partners. This workshop will be the opportunity to address top-ranked research topics in current and future materials for structures and fuel, for fission reactors and fusion reactor components. The workshop consists of four sessions. Each one will combine keynotes by leading scientists and lectures by experts in material sciences and nuclear engineering. It is held at the National Institute for Nuclear Science & Technology (INSTN) at Saclay and organized by the CEA Nuclear Energy Division (DEN) with the support of INSTN.


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MINOS workshop: program

Wednesday, December 5, 2012

8:15 • Registration

9 :00

9:05

• Welcome address, C. Gallé (MINOS), CEA Saclay (France)

• Opening address, R. Baudrillart (DEN/Dir), CEA Saclay (France)

SESSION 1: MATERIALS INNOVATION FOR NEXT GENERATION NUCLEAR ENERGY, DEC. 5TH, (a.m.)

Chairman: S. J. Zinkle, Oak Ridge National Lab. (USA) – Co-Chairman: T. D’Aletto, CEA Cadarache (France)

9:20

• Opportunities and Challenges for Materials Innovation in Nuclear Energy, S. J. Zinkle, Oak Ridge

National Lab. (USA)

10:00

• Relationship between Microstructure and Mechanical Properties in ODS Materials for Nuclear

Applications, Y. De Carlan, CEA Saclay (France)

10:25

• Innovative SiCf/SiC Composites for Nuclear Applications, L. Chaffron, CEA Saclay (France)

10:50 Morning Coffee Break

11:15

• New Textile Structures and Film Boiling Densification for SiC/SiC Components, P. David, CEA Le

Ripault (France)

11:40

• Mastery of (U,Pu)C Carbide Fuel: from Raw Materials to Final Characteristics, C. Duguay, CEA

Cadarache (France)

12:05 • The Use of Computational Thermodynamics to Predict Properties of Multicomponent Materials for

Nuclear Applications, B. Sundman, KTH (Sweden) & INSTN (France)

12:30 Lunch Break

14:00 • ANCRE Alliance: Roadmap for Nuclear Materials, F. Touboul CEA Saclay (France)

SESSION 2: IRRADIATION EFFECT, DEC. 5TH, (p.m.)

Chairman: W. J. Weber, Tennessee University (USA) – Co-Chairman: F. Willaime, CEA Saclay (France)

14:20

• Irradiation Effects in Materials for Nuclear Applications, W. J. Weber, Tennessee University

(USA)

15:00 • CEA Charged Particle Irradiation Facilities for Nuclear Material Studies, L. Beck, CEA Saclay

(France)

15:25 • Multiscale Modelling of Nuclear Fuels under Irradiation, M. Freyss, CEA Cadarache (France)

15:50 Afternoon Coffee Break

16:15 • Multiscale Modelling of Microstructure Evolution under Radiation Damage of Steels Based on

Atomistic to Mesoscale Methods, C. Domain, EDF R&D (France)

16:40

• Cluster Dynamics Modelling of Cavity and Loop Microstructure under Irradiation, T. Jourdan, CEA

Saclay (France)

17:05 • Phase and Microstructure Evolution under Irradiation : Design of Coarsening-Resistant

nanostructures, P. Bellon, Illinois University (USA)

17:30 End of the First Day


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18:30 Welcome Cocktail (La Rotonde CEA Saclay, 18h30 – 20h30)

MINOS workshop program (continue)

Thursday, December 6, 2012

SESSION 3: STRUCTURAL, MECHANICAL, CHEMICAL CHARACTERIZATION, DEC. 6TH, (a.m.)

Chairman: R. Konings, JRC-ITU & Delft University (Netherlands) – Co-Chairman: D. Féron, CEA Saclay (France)

9:00

• Nuclear Reactor Fuels: Materials with Highly Complex Behaviour, R. Konings, JRC-ITU & Delft

University (Netherlands)

9:40 • Post-Irradiation Analysis of Fission Gases in Nuclear Fuels, Ch. Valot, CEA Cadarache (France)

10:05

• Impact of Fuel Assembly Transportation on Zirconium Alloys: toward a Mechanistic

Understanding, F. Onimus, CEA Saclay (France)

10:30 Morning Coffee Break

10:55

• Materials Characterization by Tomographic Atom Probe (TAP), P. Pareige, Rouen University

(France)

11:20

• Mechanical and Thermal Resistance of Multi-Material Components for ITER, H. Burlet, CEA

Grenoble (France)

11:45

• New Characterizations at the MARS Beamline (SOLEIL Synchrotron Radiation), J.-L. Bechade, CEA

Saclay (France)

12:10 Lunch Break

SESSION 4: NUCLEAR MATERIALS AGEING, DEC. 6TH, (p.m.)

Chairman: T. shoji, Tohoku University (Japan) – Co-Chairman: B. Marini, CEA Saclay (France)

14:15

• Materials Ageing Degradation Program in Japan and Proactive Ageing Management in NPP,

T. Shoji, Tohoku University (Japan)

14:55

• The irradiation-Assisted Stress Corrosion Cracking (IASCC) issue: some Examples of Studies Carried

out at CEA, B. Tanguy, CEA Saclay (France)

15:20

• Stress Corrosion Cracking of Nickel Base Alloys in PWR Primary Water, C. Guerre, CEA Saclay

(France)

15:45

• Thermal Ageing Effects: Examples on Materials of PWR and Preventive Measures in the Design of

EPR, P. JOLY, AREVA (France)

16:10 Afternoon Coffee Break

16:35

• Plasma Facing Components: Challenges for Nuclear Materials, P. Magaud, CEA Cadarache

(France)

17:00

• Irradiation Resistance in a Fusion Environment: a Challenge for Structural Materials, J. Henry, CEA

Saclay (France)

17:25 Break

17:45 • Closing Statements, Y. Bréchet (CEA HC), CEA Saclay (France)

18:00 End of the Second Day


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MINOS workshop programme (continue)

Friday, December 7, 2012

JANNUS-SACLAY FACILITY AND LTMEX LABORATORY VISITS, DEC. 7TH, (a.m.)

8 :30 • Meeting in front of the INSTN building and departure (by bus) to the main gate of CEA Saclay

9:30 • First group to visit JANNuS-Saclay facility and LTMEx laboratory

10:30 • Transportation

10:45 • Second group to visit JANNuS-Saclay facility and LTMEx laboratory

11:45 • End of visits and departure to the Rotonde for Lunch

12:00 Lunch Break

13:30 • End of MINOS Workshop

13:45 • Departure from INSTN (bus to RER train station, « Le Guichet »)


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List of keynote and lecture speakers

• Rudy Konings, JRC-ITU & Delft University (Netherlands) • Tetsuo Shoji, Sendai University (Japan) • William J. Weber, Tennessee University (USA) • Steven J. Zinkle, Oak Ridge National Lab. (USA)

• Jean-Luc Bechade, CEA Saclay (France) • Lucile Beck, CEA Saclay (France) • Pascal Bellon, Illinois Universty (USA) • Hélène Burlet, CEA Grenoble (France) • Laurent Chaffron, CEA Saclay (France) • Frank Carré, CEA Saclay (France) • Patrick David, CEA Le Ripault (France) • Yann De Carlan, CEA Saclay (France) • Christophe Domain, EDF R&D (France) • Christelle Duguay, CEA Cadarache (France) • Michel Freyss, CEA Cadarache (France) • Catherine Guerre, CEA Saclay (France) • Jean Henry, CEA Saclay (France) • Pierre Joly, AREVA (France) • Thomas Jourdan, CEA Saclay (France) • Philippe Magaud, CEA Cadarache (France) • Jean-Paul Massoud, EDF Septen (France) • Fabien Onimus, CEA Saclay (France) • Philippe Pareige, Rouen University (France) • Bo Sundman, KTH (Sweden) & CEA-INSTN (France) • Benoît Tanguy, CEA Saclay (France) • Christophe Valot, CEA Cadarache (France)

List of local scientific committee members

• Virginie Basini, CEA Cadarache (France) • Jean-Paul Crocombette, CEA Saclay (France) • Damien Féron, CEA Saclay (France) • Christophe Gallé, CEA Saclay (France) • Bernard Marini, CEA Saclay (France) • Constantin Meis, CEA/INSTN Saclay (France) • Dominique Pêcheur, CEA Cadarache (France) • Benoît Tanguy, CEA Saclay (France)

Organization

• Christophe Gallé, CEA/MINOS and Emilie Chancrin, CEA/DEN/DMN, Saclay • Constantin Meis and Martine Mury, CEA/INSTN, Saclay

Sponsors


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Opening adress by Régis BAUDRILLART*

(*) Deputy Director of the Nuclear Energy Division (CEA/DEN, Saclay) Ladies and Gentlemen – and Dear Friends It is a pleasure for me to welcome you here today at Saclay for the first MINOS international workshop dedicated to materials science and technology for the nuclear industry. I don’t need to tell you the importance of materials expertise for nuclear energy: it is a key factor for innovation and optimization. Extending the lifetime and availability of power plants, improving safety, optimizing the fuel and cladding, developing fourth-generation nuclear systems… — materials research and innovation have a major role in all of these areas. The more we extend our knowledge of materials at the smallest scale with the objective of understanding and modeling at the same scale, the farther we will progress on the major issues facing the nuclear industry today and tomorrow. And despite the state of progress in materials research, we must pursue this work to meet present and future challenges. To address these issues, to bring together and consolidate CEA expertise, to optimize resources, and to further strengthen our relationship with our partners in the academic world, research and industry, we decided in 2011 to create the MINOS center of excellence for nuclear materials. Its ambition is to become an international reference in this field, relying both on the unique cross-divisional resources of the CEA and on its unique range of experimental and simulation tools. Some of you will have the opportunity on Friday to visit two of our facilities at Saclay:

• the Extreme Materials Technology Laboratory which designs, fabricates and characterizes innovative materials for existing and future reactors,

• and the JANNuS multi-irradiation platform which experimentally simulates the effects of irradiation on materials and quantifies changes in their microstructure.

Today MINOS has a clear roadmap assigning four major objectives:

• strengthen the scientific and technical expertise in this area,

• increase the visibility of our actions,

• foster the emergence of innovative R&D excellence,

• and promote education. For all of these reasons I am delighted that this first international workshop can contribute to these objectives by bringing together the best experts in the field. The first MINOS workshop is organized around four sessions that will provide you with opportunities for exchanges — with the highest level of expertise on materials science and nuclear engineering — on topics specifically related to materials:

• materials for future nuclear systems,

• the effects of irradiation,

• structural characterization,

• mechanical and chemical engineering,

• and aging of nuclear materials.


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As you well understand, this workshop is important for the CEA’s Nuclear Energy Division (DEN), for the CEA itself and its other research divisions, as well as for the scientific community engaged in materials for nuclear energy. By cross-fertilizing our expertise and skills, and learning from experience in France and around the world, we will be able to address the industrial challenges facing us today and tomorrow, and to benefit from scientific and technological advances. I would particularly like to thank the keynote speakers, as well as all the participants in the first MINOS international workshop. I am confident that it will lead to substantial progress in knowledge of materials. Now, as I let you begin your work, let me simply reiterate my pleasure in seeing you all here today.

MINOS Workshop, December 5, 2012


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Session 1

MATERIALS FOR NEXT GENERATION NUCLEAR ENERGY


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Opportunities and Challenges for Materials Innovation in Nuclear Energy

Steven J. ZINKLE1

1Oak Ridge National Laboratory (Oak Ridge, Tennessee, USA)

Materials performance is central to the satisfactory operation of current and future nuclear energy systems. For example, the remarkable improvement in the operation and reliability of Generation-II light water reactors (LWRs) over the past 25 years has been largely associated with improvements in steam generator materials, fuel cladding technology (composition and fabrication), and improved understanding of water chemistry impacts on corrosion and deposition. 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. The material performance demands are particularly challenging for the fuel cladding and wrapper of next-generation sodium-cooled fast reactors, where simultaneous resistance to high temperature thermal creep and radiation-induced property degradation up to doses of ~100-150 dpa is required. Several strategies can be utilized to develop structural materials with simultaneous high radiation resistance, high strength, good toughness and corrosion resistance, and moderate fabrication cost. There are three general approaches for designing radiation resistance: Nanoscale precipitates or interfaces to produce high point defect sink strength (e.g., oxide dispersion strengthened (ODS) and next-generation ferritic/martensitic steels with high particle densities); purposeful utilization of immobile vacancies (e.g., SiC/SiC ceramic composites); and utilization of radiation-resilient matrix phases (e.g., ferritic instead of austenitic steel matrix, etc.). High-performance steels designed using computational thermodynamics are demonstrating promising capability to produce a high density of highly stable nanoscale precipitates that could serve as efficient point defect recombination centers during irradiation, and also provide good thermal creep strength at high temperatures [1]. Figure 1 shows an example of improvements in high temperature thermal creep properties for a 9%Cr-1%Mo ferritic martensitic steel that was achieved simply by slightly altering the thermomechanical processing procedure. Figure 2 compares the fracture toughness behavior of an advanced ODS ferritic steel before and after low dose neutron irradiation at 300oC [2]. The ductile to brittle transition temperature (DBTT) in the LT orientation remained below -150oC with a shift in the DBTT of about 12oC; the corresponding shift in the DBTT for EUROFER97 (9Cr-2WVTa) ferritic/martensitic steel was about 39oC. Higher dose studies are in progress. It will be important for nuclear energy researchers to continue to closely interact with the broader materials science and engineering community in order to effectively leverage innovations that continue to occur in the broad field of materials science. For example, practical aspects used in the aerospace industry to reduce the time from invention of a new alloy system to code qualification and commercialization could be useful for development of new structural materials for nuclear energy systems. In the future, utilization of emerging advanced manufacturing processes such as additive manufacturing 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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Following the accident at the Fukushima Daiichi site in Japan, there is increasing interest in exploring accident tolerant or enhanced safety margin fuel systems for existing and future reactors, which could potentially provide increased response time or reduced consequences via reduced enthalpy production, reduced hydrogen production, and delayed clad rupture or fission product release during a loss of coolant accident compared to conventional Zr alloy cladding/ monolithic UO2 fuel systems. Although all alternative fuel systems have technological or neutronic shortcomings, exploratory research would be useful to quantify their potential improved accident tolerance so that an informed decision on the best option(s) for future LWR fuel systems can be reached. If research results on a particular accident tolerant concept prove to be promising, it might be possible to initiate confirmatory tests in commercial reactors within about 10 years.

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8090

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200

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100 1000 10000 100000

Comparison of Creep Rupture Behavior of 9Cr Steels

Klueh et al. (2007)Hollner et al. (2010)

Str

ess (

MP

a)

Rupture Life (h)

F82H N&T

Mod 9Cr-1Mo N&T

Mod 9Cr-1Mo TMT

650 oC

Fig. 1: Comparison of creep rupture behavior

of 9%Cr steels at 650oC after conventional and new thermomechanical treatment [1].

Fig. 2 : Comparison of the fracture toughness of 14YWT ODS ferritic steel before and after neutron irradiation to 1.5 dpa at 300oC [2].

References [1] S.J. Zinkle, N.M. Ghoniem. J. Nuclear Materials 417 (2011) 2. [2] D.A. McClintock, M.A. Sokolov, D.T. Hoelzer, R.K. Nanstad.J. Nuclear Materials 392 (2009) 353.


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Relationship between Microstructure and Mechanical Properties in ODS Materials for Nuclear Application

Yann De CARLAN1

1CEA-DEN-DMN, Service de Recherches Métallurgiques Appliquées, SRMA (Saclay, France)

Oxide Dispersion Strengthened ferritic/martensitic alloys are developed as prospective cladding materials for future Sodium-Cooled-Fast-Reactors (GEN IV) [1]. These advanced alloys present a good resistance to irradiation and a high creep rupture strength due to a reinforcement by the homogeneous dispersion of hard nano-sized particles (such as Y2O3 or YTiO). ODS alloys are elaborated by powder metallurgy, consolidated by hot extrusion and manufactured into cladding tube using the pilger cold-rolling process [2, 3]. ODS alloys present usually low ductility and high hardness. The aim of this talk is to present the specificity of the metallurgy of ODS materials in relationship with the main mechanical properties (tensile and creep properties, toughness, transition temperature…). Two types of alloys will be presented: Fe-9Cr martensitic ODS and Fe-14Cr ferritic ODS alloys. Mechanical properties of the materials depend on the metallurgical state (fine grains, recrystallized, martensitic) and very different behaviors are observed as a function of final microstructure. For example, for a Fe-9CrODS alloy, tempered martensite lets obtaining material with high strength

whereas softened ferrite see figure 1 [4] tolerates high deformation levels.

Tempered martensitic structure

Softened ferritic structure

Fig. 1: Electron Backscatter Diffraction maps for a Fe-9Cr ODS associated with the tensile properties at room temperature.


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References [1] P. Dubuisson, Y. de Carlan, V. Garat, M. Blat, ODS Ferritic/martensitic alloys for Sodium Fast

Reactor fuel pin cladding. Journal of Nuclear Materials 428 (2012) 6–12. [2] T. Narita, S. Ukai, T. Kaito, S. Ohtsuka, T. Kobayashi, Development of Two-Step Softening Heat

Treatment for Manufacturing 12Cr–ODS Ferritic Steel Tubes. Journal of Nuclear Science and Technology, Vol. 41, No. 10, p. 1008–1012 (October 2004).

[3] S. Ohtsuka, S. Ukai, M. Fujiwara, T. Kaito, T. Narita, Improvement of 9Cr-ODS martensitic steel properties by controlling excess oxygen and titanium contents. Journal of Nuclear Materials 329-333(1): 372-376.

[4] L. Toualbi, C. Cayron, P. Olier, R. Logé, Y. de Carlan. Relationships between mechanical behavior and microstructural evolutions in Fe 9Cr-ODS during the fabrication route of SFR cladding tubes. Submitted in Journal of Nuclear Materials, April 2012.


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Innovative SiC/SiC Composite for Nuclear Applications

Laurent CHAFFRON1, Cédric SAUDER1, Christophe LORRETTE1, Laurent BRIOTTET2,

Aurore MICHAUX1, Lionel GÉLÉBART1, Aurélie COUPÉ1, Maxime ZABIEGO3,

Marion LE FLEM1, Jean-Louis SÉRAN1

1CEA-DEN-DMN, Service de Recherche Métallurgiques Appliquées, SRMA (Saclay, France)

2 CEA-DRT, Laboratoires d’Innovation pour les Technologies des Energies, LITEN (Grenoble, France)

3 CEA-DEN-DEC, Service d’Etudes et de Simulation du Comportement des Combustibles, SESC (Cadarache, France)

Among various refractory materials, SiC/SiC ceramic matrix composites (CMC) are of prime interest

for fusion and advanced fission energy applications, due to their excellent irradiation tolerance and

safety features (low activation, low tritium permeability,…). Initially developed as fuel cladding

materials for the Fourth generation Gas cooled Fast Reactor (GFR), this material has been recently

envisaged by CEA for different core structures of Sodium Fast Reactor (SFR) which combines fast

neutrons and high temperature (500°C). Regarding fuel cladding generic application, in the case of

GFR, the first challenge facing this project is to demonstrate the feasibility of a fuel operating under

very harsh conditions that are (i) temperatures of structures up to 700°C in nominal and over 1600°C

in accidental conditions, (ii) irradiation damage higher than 60 dpaSiC, (iii) neutronic transparency,

which disqualifies conventional refractory metals as structural core materials, (iv) mechanical

behavior that guarantees in most circumstances the integrity of the first barrier (e.g.: > 0.5%), which

excludes monolithic ceramics and therefore encourages the development of new types of fibrous

composites SiC/SiC adapted to the fast reactor conditions. No existing material being capable to

match all these requirements, CEA has launched an ambitious program of development of an

advanced material satisfying the specifications [1]. This project, that implies many laboratories, inside

and outside CEA, has permitted to obtain a very high quality compound that meets most of the

challenging requirements. We present hereinafter few recent results obtained regarding the

development of the composite. One of the most relevant challenges was to make a gastight compo-

site up to the ultimate rupture. Indeed, multicraking of the matrix is the counterpart of the damageable

behavior observed in these amazing compounds. Among different solutions envisaged, an innovative

one has been successful. It consists of inserting a metallic layer between two tubes of CMC [2]. The

concept, illustrated in figure 1, guaranties a perfect helium tightness up to fracture of the CMC.

Fig. 1: Sandwich cladding concept: tightness is ensured up

to CMC failure thanks to the elastic metallic layer.

Fig. 2: Sandwich cladding Cross

section (metal is in white).


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Another challenge was to prepare a representative cladding with very strict geometrical tolerances.

Revisiting the fabrication of the entire breading process has allowed to ensure a perfect geometry of

the final tube. Thanks to the high quality of manufacture and the high level of purity of composite

materials manufactured at CEA, few tens of CMC objects (tubes, disks and plates) have been

prepared in order to be irradiated in the Russian reactor “BOR 60”. For the first time, composite

materials will be submitted to swift neutrons at very high damaging doses (up to 80 dpaSiC) between

400 and 520°C. Post irradiation examinations expected for 2015 should give reliable results on the

behavior of this multi-materials component. In parallel, other basic researches are conducted to

improve the properties of the CMC and round off the understanding [3, 4, 5]. Some new results

allowed to extend the field of use of the CMC through an optimization of the interphase of the

composite. The figure 4 shows the relative elongation of a CMC after a two hours dwell time

annealing in argon at different temperatures: optimized composite can sustain very high temperature

without drastic drop of its mechanical properties.

Fig. 3: CMC specimen prepared

for BOR60 irradiation

Fig. 4: Evolution of the relative elongation of two

composites with the annealing temperature: optimized

CVI conditions to improve mechanical properties.

References [1] L. Chaffron, J. L. Séran, C. Sauder, C. Lorrette, A. Michaux, L. Gélébar1, A. Coupé, SiC/SiC

Composite Materials for Fast Reactor Applications. Proceedings of ICAPP 2011, Nice, France, May 2-5, 2011, Paper 11433.

[2] M. Zabiégo, C. Sauder, C. Lorrette, P. Guédeney, Tube multicouche amélioré en matériau composite à matrice céramique, gaine de combustible nucléaire en résultant et procédés de fabrication associés.Patent submitted 1 August 2011, in French.

[3] C. Sauder, J. Lamon, Influence of fiber surface roughness on mechanical behaviour of SiC/SiC minicomposites with Hi-Nicalon S and SA3 reinforcement. 35ème International Congress on Advanced Ceramic and Composites, Daytona beach 25 Janvier 2011.

[4] E. Buet, C. Sauder, S. Poissonnet, P. Brender, R. Gadiou, C. Vix-Guterl, Influence of chemical and physical properties of the last generation of silicon carbide fibres on the mechanical behaviour of SiC/SiC composite. Journal of the European Ceramic Society, 2012. 32(3): p. 547-557.

[5] A. Coupé, H. Maskrot, E. Buet, A. Renault, P.J. Fontaine, L. Chaffron, Dispersion Behavior of Laser-synthesized silicon carbide nanopowders in ethanol for Electrophoretic Infiltration. Journal of the European Ceramic Society, Vol 32, Issue 14, 3837-3850, 2012.


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New Textile Structures and Film-Boiling Densification for SiC/SiC Components

Patrick DAVID1

, Joelle BLEIN1, Yannick PIERRE1, Denis ROCHAIS1, Maxime ZABIEGO2

1CEA-DAM, Département Matériaux (Monts, France)

2CEA-DEN-DEC, Service d’Etudes et de Simulation du Comportement des Combustibles, SESC (Cadarache, France)

Among all ceramic candidates, silicon carbide (SiC) materials present the best properties for use in very harsh nuclear environments. Indeed, they possess an excellent behavior under high neutron irradiation, high thermal conductivity and mechanical properties at high temperature in addition to a chemical inertness. Meanwhile, as monolithic ceramics are too brittle, it is necessary to use SiC in the form of SiC/SiC materials, which exhibit a high fracture toughness and allow the manufacturing of both very thin and large pieces. Principal studies of nuclear applications have been conducted since the middle of the 1990’s, for blanket of wall fusion reactors [1]. SiC/SiC composites have also been recently envisaged as alternatives to zircaloy fuel claddings in order to cope with core overheating accidents of Light Water Reactors [2]. During the last decade, intensive investigations on 4th generation fission reactor applications have been conducted [3], mainly for fuel claddings, and also, more recently, for other components, such as the Hexagonal Tube (HT) for pin assembling, control rods, heat exchangers or thermal barriers for the Hot Gas Duct. Several developments and optimizations are necessary to meet the specifications of structural components in nuclear systems which differ strongly from those of common aeronautical and spatial applications. The performed studies have addressed several technological points that were considered as bottlenecks. For Gas cooled Fast Reactor claddings, new textile processes have been developed for manufacturing of honeycomb structures [4] or closed tubular braids [5] (Figure 1), used as fibrous reinforcement structures. Control of the geometry can be improved using graphite tools, during densification, and by addition of a sacrificial layer that is machined away after densification (Table 1), (Figure 2). The internal surface smoothness has been increased through an initial Chemical Vapor Deposition step on the graphite rod used as a mandrel for pin braiding. As SiC/SiC materials possess a limited tolerance when it comes to the set deformation, a flexible ceramic porous bond, constituted of a ceramic textile structure, has been proposed [6] and characterized to improve the fuel pellet-clad interaction. Concerning the Hexagonal Tube, studies have been performed on the very particular film-boiling process (Figures 3 and 4). The aim was to reduce the important densification time and cost, due to the quite large dimensions and thickness of this component. Depending on the chemical precursors used, the main difficulty consisted in controlling either the matrix composition, which can contain excess carbon, or the microstructure, which can be less ordered and conductive than CVI deposits. A few investigations have also been initiated on SiC/SiC with low density and conductivity, to manufacture a thermal barrier for the Hot Gas Duct.

Owing to the stringent specifications set for nuclear components, the development of SiC/SiC materials, and more particularly for cladding applications, is an ambitious target, both from technological and scientific standpoints. Complementary tests and characterizations are still necessary to prove that these materials are consistent with the targeted performance. All the knowledge and knowhow developed during these studies should be useful to obtain technological solutions for tailored and reliable SiC/SiC components for high-temperature (nuclear) applications.


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Table 1: Accuracies obtained for the geometrical dimensions of a 10-cm long, 4-layered bi-axial braid.

External Diameter (mm) General straightness (mm) Local straightness (mm)

[8.96, 8.99] < 0.03 0.04

References

[1] L.L. Snead, R.H. Jones, A. Kohyama, P. Fenici, Status of silicon carbide composites for fusion. J. Nucl. Mater. 233–237, 26 (1996).

[2] H. Feinroth, B. Hao, L. Fehrenbacher, Progress in developing an impermeable, high temperature ceramic composite for advanced reactor clad and structural applications. ICAPP Proceedings (2002).

[3] W. Corwin, The Gas Fast Reactor (GFR) Survey of Material Experience and R&D Needs to Assess Viability. Ref: ONRL/TM-2004/99 (2004).

[4] CEA Patent, CEA Patent, Method for producing a cellular fibrous structure, P. David. WO 2009/0065794, 28/05/2009.

[5] CEA Patent, Architecture fibreuse tubulaire fermée et procédé de fabrication, P. David, JL Bonnand, B. Bompard, PCT/EP2010/067736, 18/11/2010.

[6] CEA Patent, Joint d’interface solide à porosité ouverte pour crayon de combustible nucléaire et pour barre de commande nucléaire. M. Zabiego, P. David, A. Ravenet, D. Rochais, PCT/EP2011/060001, 16/06/2011.

(1-a) (1-b)

Fig. 1: The closed extremity of the SiC/SiC densified braid (1-a) and a representation

(simulation) of the position of the yarns (1-b).

Fig. 4: The microstucture of a SiC/SiC composite (hexagonal tube application)

densified with the film-boiling technique.

Fig. 2: A view of the surface of the SiC/SiC densified braid,

after machining (Ra ~ 10 m).

Fig. 3: A view of the film-boiling reactor. The fibrous structure is immersed in a liquid precursor and brought to about 1000°C

through radio-frequency heating.


21

Mastery of (U,Pu)C Carbide Fuel: from Raw Materials to Final Characteristics

Christelle DUGUAY1

1CEA-DEN-DEC, Service Plutonium Uranium et Actinides Mineurs, SPUA (Cadarache, France)

Mixed uranium plutonium carbide is for many years an advanced and alternative fuel to the mixed oxide being developed for Gas and Sodium Fast Reactors due to its high thermal conductivity and heavy-metal density [1].

Compared to that of oxide fuels, the fabrication of carbide fuels meeting all required specifications (among which are an oxygen content lower than 1000 ppm and a predominant open porosity Po/Pt>50%) is a much more difficult and challenging task. Due to their extreme reactivity with oxygen and moisture, fabrication and handling of carbide fuels are indeed performed in gloveboxes maintained at purity levels needed to provide operational safety and to obtain high-quality fuels, under a dynamic flow of nitrogen (O2 and H2O, each less than 50 ppm). Nevertheless, oxygen pick up during processing of the material seems unavoidable, even at room temperature, and must be thus limited by suitable procedures. A significant loss of plutonium may also be caused by vaporization during the heating steps, especially at high temperatures. Such a loss must be limited for radiological reasons (the evaporated plutonium condenses in the cold parts of furnaces) and since it affects the composition of the carbide. As shown on the simplified flow sheet presented in fig.1, the main steps in the fabrication of carbide fuel pellets are as follows:

- Vacuum carbothermic synthesis of carbide fuel in the temperature range of 1450-1650°C from a blend of UO2, PuO2 and graphite powders pressed into compacts,

- Crushing and milling of compacted samples, - Consolidation of carbide powders into fuel pellets by cold pressing and sintering at 1750°C

under Ar + 5% H2 gas, To achieve specified pellet density (80 %TD) and porosity, a pore former (such as zinc stearate) can be added to the fuel powder.

PuO2 UO 2 C

PRESSAGE

+

FRITTAGE

MELANGE

BROYAGE é

PuO2 UO 2 C

COMPACTING CARBOREDUCTION

CRUSHING +

GRINDING

BINDER + PORE FORMER

SINTERING

BLENDING

BLENDING

CONTROLS

UPuC SYNTHESIS

PELLETIZING DENSIFICATION

PRESSING


22

Fig. 1: Mixed carbide pellets fabrication flowsheet. The sintered carbides are mainly composed of a monocarbide phase (U,Pu)C, with traces of a sesquicarbure phase (U,Pu)2C3. Their density reaches 75% (when pore former is added) to 91% of theoretical density. They contain residual oxygen as impurity in varying amounts, depending on the fabrication conditions, and above the upper specified limit. The residual oxygen content decreases when the open porosity increases (fig. 2). For an open porosity higher than 40%, relatively low levels of oxygen, between 1000 and 3000 ppm, can be obtained, even if the raw carbide powder has a high oxygen content (sintering classically leads to a significant reduction in the oxygen content).

Fig. 2: Carbides oxygen content as a function of their open porosity.

The use of new oxide precursors, whose synthesis is based on the co-conversion of actinides [2], may simplify the manufacturing process by reducing the number of process steps, and may avoid the additional oxygen contamination of the final product as well as the radiation exposure of the operators. Faster reaction rates during carbothermic reduction could be achieved, as well as complete conversion of oxides to carbide at relatively low temperatures so as to keep plutonium volatilization at a reasonably low level. So, (U,Pu)O2 or even precursors obtained by direct incorporation of carbon graphite or a carbon molecule in the structure of the mixed oxalate could be used as starting materials instead of UO2 and PuO2. References [1] Hj. Matzke, Science of advanced LMFBR fuels. A Monograph on sold State Physics, Chemistry

and Technology of Carbides, Nitrites and Carbonitrides of Uranium and Plutonium. Elsevier Science & Technology Books, 1986.

[2] A. Handshuh, Synthèse des carbures d’U-Pu : influence des précurseurs et mécanismes réactionnels de la carbothermie des systèmes lanthanides et actinides. These de Doctorat de l’Université des Sciences et technologies de Lille, octobre 2010.


23

The Use of Computational Thermodynamics to Predict Properties of Multicomponent Materials for Nuclear Applications

Bo SUNDMAN1, Christine GUÉNEAU2

1KTH (Stockholm, Sweden), ISNTN (Saclay, France)

2CEA-DEN-DPC, Service de la Corrosion et du Comportement des Matériaux dans leur Environnement, SCCME (Saclay, France)

Computational Thermodynamics is based on physically realistic models to describe metallic and oxide crystalline phases as well as the liquid and the gas in a consistent manner [1]. The models are used to assess experimental and theoretical data for many different materials and several thermodynamic databases has been developed for steels, ceramics, semiconductor materials as well as materials for nuclear applications. Within CEA a long term work is ongoing to develop a database for the properties of nuclear fuels and structural materials [2]. An overview of the modeling technique will be given and several examples of the application of the database to different problems, both for traditional phase diagram calculations and its use in simulating phase transformations. The following diagrams (Fig. 1, Fig. 2 and Fig.3) show calculations in the U-Pu-O system.

Fig. 1 : U-Pu-O phase diagram at 1273°K.

Fig. 2 : U-Pu-O phase diagram at 673°K showing the miscibility gap in the MO2 phase together

with experimental data.


24

Fig. 3 : Oxygen chemical potentials at different temperatures and varying oxygen content for systems with Pu/u ratios around 0.3, together with experimantal data.

References [1] H. Lukas, S. Fries, B. Sundman, Computational Thermodynamics. Cambridge University Press

(2007). [2] C. Guéneau, N. Dupin, B. Sundman, C. Martial, J.-C. Dumas, S. Gossé, S. Chatain, F. De

Bruycker, D. Manar, R.J.M. Konings, Thermodynamic modeling of advanced oxide and carbide nuclear fuels: description of the U-Pu-O-C systems. J. Nucl Mater. 419 (2011) 145


25

ANCRE Alliance: Roadmap for Nuclear Materials

Françoise TOUBOUL1, Frank CARRÉ2

1CEA-DEN-DISN, Scientific and Technological Basic Research Program (Saclay, France)

2CEA-DEN, Scientific Direction (Saclay, France)

Created in 2009 by the Higher Education and Research ministry and by the Ecology ministry, ANCRE, the National Alliance for Energy Research Coordination aims at enhancing the efficiency of French research in the field of energy by promoting partnerships and synergies between public and private sectors (research organizations, universities and companies). ANCRE aims to propose a coordinated strategy for research and innovation projects. Beyond its four founding members, CEA, CNRS, IFPEN and CPU, ANCRE brings together all the French public research organizations concerned with energy issues, and has strong links with the industrial sector. Among the 10 programmatic groups of ANCRE, one is specifically dedicated to Nuclear Energies (fission and fusion). This group has proposed roadmaps in five scientific fields, considered as strategic for R&D, in relation to industrial objectives and scientific bottlenecks: nuclear materials, nuclear chemistry, reactor physics, instrumentation and fusion. For twenty to thirty years, R&D on nuclear materials has evolved from the heavy metallurgy of the first generation of power reactors to the nano-materials science under extreme conditions for present and future needs. Nuclear systems are characterized by extreme operating conditions: high temperatures, mechanical stresses, radiations, corrosive environment, and long durations. In order to deal with these extreme conditions, it is necessary to have a sound knowledge of the materials, to the finest scale. R&D development was made possible by advances in materials science, in relation to more efficient observation means (now reaching the atom scale) and deeper control of the microstructure. Development of simulation methods at the atomic level (ab initio, classical molecular dynamics, kinetic Monte Carlo, etc.) have also allowed a better understanding of phenomena at their most fundamental level. Material performance issues, however, remain significant, as the performance targets have changed: more severe operating conditions (higher neutron fluence and very high temperature), and new objectives for reliability and security, in order to prevent and safely manage external aggressions (earthquake, aircraft impact, etc.) and severe internal accidents. Manufacturers and researchers must rely on robust methods to assess and justify the lifetime of existing nuclear reactors, and to develop new materials (incrementally or leapfrogging) for current and future nuclear systems. Priority objectives have been defined by ANCRE members and subsequent scientific locks have been determined, leading to program proposals: material for reactor vessel or internal circuits, corrosion mechanisms, zirconium alloys, fuel materials, nano-reinforced steels, surface engineering, advanced metallurgical processes, composite materials, materials for fusion, refractory materials, containment materials (concrete, glass, clay), structural mechanics, multi-scale modeling and simulation, experiments. At the same time they recommend directions for cooperative research, ANCRE roadmaps also identify existing laboratories able to contribute to these research goals, and suggest new research clusters or structures when appropriate, as well as other frameworks of cooperation at national, European or international level. ANCRE also recommends initiatives to tighten links between research and education.


26

ANCRE roadmap on materials for nuclear energies fully concurs that multi-scale and multi-physics modeling and simulation tools based on the most accurate knowledge of physicochemical phenomena at various scales, together with characterization tools at the same scales, are of crucial importance for predicting materials performance and lifetime in service conditions, and for guiding research on materials for future nuclear systems.


27

Session 2

IRRADIATION EFFECTS


28


29

Irradiation Effects in Materials for Nuclear Applications

William J. WEBER1

1Department of Materials Science & Engineering, University of Tennessee (Knoxville, TN, USA)

There has been a resurgence of international interest in nuclear energy as a clean energy source, and the continued success of existing nuclear reactors, the promise of advanced reactor designs, and the acceptance of used nuclear fuel disposition options are dependent on the performance of materials in extreme nuclear environments. The effects of irradiation on materials properties and performance are critical to safe and reliable reactor operations, efficient use of uranium resources, reduced production of nuclear waste, and acceptable used nuclear fuel disposition [1]. This lecture will provide an introduction to irradiation effects in materials for nuclear applications. The fundamental irradiation damage mechanisms will be described, and the effects of different irradiation environments for nuclear fuels (nuclear fission damage), structural components (neutron damage), and used nuclear fuel and nuclear waste forms (alpha decay damage) will be discussed.

Irradiation damage in materials for nuclear applications primarily results from the production of energetic particles in fission, nuclear reaction and radioactive decay events. The interaction of these energetic particles (fission products, fast neutrons, protons, alphas, and recoil nuclei) with materials results in the production of atomic-scale defects from ballistic collisions and introduction of new chemical elements. In the case of high-energy fission products, the intense ionization along the fission product path can also introduce defects or damage [2]. These irradiation damage processes control the long-term evolution of the materials response to the production and diffusion of defects, fission products and nuclear reaction products. With the exception of fast neutrons, the energetic particles produced in nuclear environments are charged ions that can be also be produced using ion accelerator facilities, such as the JANNuS facility in France [3] or facilities in the USA [4, 5]. Even in the case of a fast neutron, 50 to 100 keV of kinetic energy is transferred on average to a primary knock-on atom (PKA) [6], which is essentially an energetic ion that produces a cascade of displaced atoms through screened-Coulomb collisions, as illustrated in Fig. 1 for silicon carbide [7].

Ion-beam irradiation techniques can be effective in simulating neutron irradiation effects [6, 8], fission damage [9] and alpha-decay damage [10] on short laboratory time scales over a large range of experimental conditions in order to develop more detailed scientific understanding and predictive models of the complex evolution microstructure and phase changes under irradiation. The unique effects of ion irradiation and the use of ions to study fast neutron irradiation effects is a critical topic, since development and qualification of new structural materials requires neutron doses that are too high to be obtained in existing nuclear test reactors or spallation neutron sources.

The advantages and disadvantages of using ions to simulate nuclear radiation environments will be reviewed, with specific examples provided. The integration of computer simulations into these irradiation studies have advanced the interpretation of experimental results

Fig. 1: Defects created by a 50 keV Si PKA cascade in silicon carbide (only

defects are shown) [7].


30

[2, 7] and provided a more comprehensive atomic-level understanding and predictive models of irradiation damage processes. References [1] Y. Guerin, G. S. Was, S. J. Zinkle, Materials Challenges for Advanced Nuclear Energy Systems.

MRS Bulletin 34 [1] : 10-14 (2009). [2]R. Devanathan, W. J. Weber, Simulation of collision cascades and thermal spikes in ceramic. Nucl.

Instrum. and Methods Phys. Res. B 268 : 2857-2862 (2010). [3] Y. Serruys, M.-O. Ruault, P. Trocellier, S. Henry, O. Kaitasov, Ph. Trouslard, Multiple ion beam

irradiation and implantation : JANNUS project. Nucl. Instrum. and Methods Phys. Res. B 240 : 124-127 (2005).

[4] I. O. Usov, D. J. Devlin, J. Won, A. Kossoy, J. A. Valdez, Y. Q. Wang, K. E. Sickafus, Medium energy ion irradiation capability for studies of radiation damage effects over a wide temperature range. Nucl. Instrum. and Methods Phys. Res. B 269 : 2734-2739 (2011).

[5] S. Thevuthasan, C. H. F. Peden, M. H. Engelhard, D. R. Baer, G. S. Herman, W. Jiang, Y. Liang, W. J. Weber, The Ion Beam Materials Analysis Laboratory at the Environmental Molecular Sciences Laboratory. Nucl. Instrum. and Methods in Phys. Res. A 420 [1-2]: 81-89 (1999).

[6] G. S. Was, Fundamentals of Radiation Materials Science (Springer, Heidelberg, 2007). [7] F. Gao, W. J. Weber, Atomic-level study of ion-induced nanoscale disordered domains in silicon

carbide. Applied Physics Letters 82 [6] : 913-915 (2003). [8] T. R. Allen, J. Gan, J. I. Cole, M. K. Miller, J. T. Busby, S. Shutthanandan, S. Thevuthasan,

Radiation response of a 9 chromium oxide dispersion strengthen steel to heavy ion irradiation. J. Nuclear Materials 375 : 26-37 (2008).

[9] W. Li, M. Lang, A. J. W. Gleadow, M. V. Zdorovets, R. C. Ewing, Thermal annealing of unetched fission tracks in apatite. Earth and Planetary Science Letters 321-322 : 121-127 (2012).

[10] S. Moll, G. Sattonnay, L. Thomé, J. Jagielski, C. Decorse, P. Simon, I. Monnet, W. J. Weber, Irradiation damage in Gd2Ti2O7 single crystals: Ballistic versus ionization processes, Physical Review B 84 [6]: 064115 (2011).


31

CEA Charged Particle Irradiation Facilities for Nuclear Material Studies

Lucile BECK1

1CEA-DEN-DMN, JANNUS Saclay laboratory (Saclay, FRANCE)

To study the effects of radiation on nuclear materials, the CEA uses various irradiation facilities producing neutrons or charges particles. Neutron facilities have the great advantage to directly produce neutron damages, but have also some drawbacks. The samples are activated and consequently their characterizations need dedicated hot cells, the reactors have fixed experimental parameters and the experiments are long and costly. In order to better understand the mechanisms of radiation damage, experimental simulation can be also conducted with charged particles. In this case, experimental irradiation conditions (temperature, dose, flux, energy) are well controlled and the irradiated samples can be characterized with conventional analytical methods. This presentation will describe the CEA facilities devoted to ion or electron irradiations for material studies. Examples of irradiations performed at JANNUS (Joint Accelerators for Nano-science and NUclear Simulation) will be detailed (Fig. 1).

Fig. 1: Schematic view of the JANNUS platform at Saclay.


32


33

Multiscale Modelling of Nuclear Fuels under Irradiation

Michel FREYSS1

1CEA-DEN-DEC, Service d’Etudes et de Simulation du Comportement des Combustibles, SESC (Cadarache, France)

The fuel element under irradiation is submitted to a wide variety of coupled phenomena involving among others temperature, mechanical load, radiation damage, chemical interaction between the material and the fission products. The PLEIADES fuel performance software environment [1,2] developed by CEA can predict the behaviour of standard or innovative fuel elements under operating conditions. It is nevertheless still a challenge for R&D to refine the laws used in fuel performance codes by a more physically based description of the fuel materials, and improve both the understanding of the phenomena involved during irradiation and the capability to predict the fuel behaviour. This goal requires to decorrelate the complex phenomena involved in the material evolution by conducting studies towards the atomistic level. It also requires to couple post-irradiation examinations (PIE) with separate-effect experiments and various modeling approaches at the relevant scales. In particular, basic research on fuel materials focuses on the evolution under irradiation of the microstructure, the transport properties of defects, fission products, helium, as well as their thermochemistry.

Fig. 1: Multiscale scheme applied to oxide fuels, as implemented in the European F-BRIDGE project [3].

First

Principles

Empirical potential / classical molecular dynamics

nm µm mm m

distance

ps

µs

s

Days / years

time

ms

Thermo-mechanical Finite Element Modeling Fuel performance codes

atoms Grain(s) Pellet(s)

Thermodynamics

modeling

Kinetic models/ Rate theory Cluster dynamics

Homogenization

micro/macro


34

The overall strategy for a multiscale modeling scheme of nuclear fuels will be presented, with some examples of applications making use of atomic scale approaches: First-principles electronic structure calculations used to get insight into the atomic transport properties of point defects [4] and classical molecular dynamics used to model the ballistic damage created by the recoil of fission products [5]. The importance of coupling the multiscale modeling approach to experimental studies will also be emphasized by some examples to illustrate the characterization of the oxygen diffusion [6] and the bubble formation of fission gases [7] in UO2. Such separate effects experiments should be seen as guides to orientate modeling, but also as essential tools to validate the approximations of the modeling methods. References [1] B. Michel, C. Nonon, J. Sercombe, F. Michel, V. Marelle. to appear in Nuclear Technology (2012). [2] L. Noirot. Nuclear Engineering and Design 241, 2099 (2011). [3] F-BRIDGE project (Basic Research for Innovative Fuel Design for GEN IV systems), project of the

Seventh Framework Programme of the European Commission, http://www.f-bridge.eu/ [4] B. Dorado, D. A. Andersson, C. R. Stanek, M. Bertolus, B. P. Uberuaga, G. Martin, M. Freyss, P.

Garcia. Phys. Rev. B 86, 035110 (2012). [5] G. Martin, P. Garcia, L. Van Brutzel, B. Dorado, S. Maillard, Nuclear Instruments and Methods.

Physics Research B 269, 1727 (2011). [6] B. Dorado, P. Garcia, M. Freyss, G. Carlot, M. Fraczkiewicz, B. Pasquet, G. Baldinozzi, D.

Simeone, M. Freyss, M. Bertolus. Phys. Rev. B 83, 035126 (2011). [7] P. Garcia, G. Martin, C. Sabathier, G. Carlot, A. Michel, P. Martin, B. Dorado, M. Freyss, M.

Bertolus, R. Skorek, J. Noirot, L. Noirot, O. Kaitasov, S. Maillard. Nuclear Instruments and Methods. Physics Research B 277, 98 (2012).

http://www.f-bridge.eu/


35

Multiscale Modelling of Microstructure Evolution under Radiation Damage of Steels Based on Atomistic to Mesoscale Methods

Christophe DOMAIN1, 2

1 EDF R&D - MMC (Moret sur Loing, France)

2 EDF-CNRS joint laboratory EM2VM (Study and Modeling of the Microstructure for Ageing of Materials)

Structural metallic materials used in nuclear facilities are submitted to irradiation which induce the creation of large amounts of point defects, which leads to modifications of the microstructure and the mechanical properties. In nuclear power plants, the main structural materials are: the pressure vessel (ferritic steels), the internal structure (austenitic steels). In order to simulate the microstructure evolution with the objective to predict it, multiscale modelling tools are developed (Fig. 1). For this purpose different simulation methods are used and developed in order to treat the different physical phenomena occurring at different time scales and length scales: ab initio, classical molecular dynamics, kinetic Monte Carlo, dislocation dynamics, phase field [1]. These simulations are very CPU demanding and take advantage of the development of High Performance Computing machines.

Finite elements

ab initio

Molecular dynamics

Mesoscopic

Multi-scale

modelling

1nm3

0 - psns(10-30nm)3

cm3

µm3

h-year

s - h

(30-100nm)3

m3

40 years

Micro-macro

Dislocationdynamics

Finite elements

ab initio

Molecular dynamics

Mesoscopic

Multi-scale

modelling

1nm3

0 - psns(10-30nm)3

cm3

µm3

h-year

s - h

(30-100nm)3

m3

40 years

Micro-macro

Dislocationdynamics


36

Fig. 1: Multiscale modelling scheme applied within the PERFORM-60 project to the pressure vessel and internal material microstructure.

The microstructure evolution under irradiation is obtained starting from the neutron spectrum to obtain the primary damage (displacement cascades), followed by the evolution of the point defects formed and their accumulation (Fig. 2).

Spectre de neutron

1.E+08

1.E+09

1.E+10

1.E+11

1.E+12

1.E-08 1.E-06 1.E-04 1.E-02 1.E+0

0

1.E+0

2

Flu

x (

n/c

m2/s

)

1.E-08

1.E-05

1.E-02

1.E+01

1.E+04

1E-05 1E-04 1E-03 1E-02 1E-01 1E+00 1E+01

EPKA (MeV)

PKA F

lux

(PKA/µ

m3/M

eV/s

)

PWR

Neutron

spectrum

PKA

spectrum

Primary

damage

Short term

evolution

Interactions

defects -

dislocations

Exper.

resolvable

defects

Spectre de neutron

1.E+08

1.E+09

1.E+10

1.E+11

1.E+12

1.E-08 1.E-06 1.E-04 1.E-02 1.E+0

0

1.E+0

2

Flu

x (

n/c

m2/s

)

1.E-08

1.E-05

1.E-02

1.E+01

1.E+04

1E-05 1E-04 1E-03 1E-02 1E-01 1E+00 1E+01

EPKA (MeV)

PKA F

lux

(PKA/µ

m3/M

eV/s

)

PWR

1.E-08

1.E-05

1.E-02

1.E+01

1.E+04

1E-05 1E-04 1E-03 1E-02 1E-01 1E+00 1E+01

EPKA (MeV)

PKA F

lux

(PKA/µ

m3/M

eV/s

)

PWR

Neutron

spectrum

PKA

spectrum

Primary

damage

Short term

evolution

Interactions

defects -

dislocations

Exper.

resolvable

defects

Fig. 2: Microstructure modelling under irradiation. The point defects created (vacancy and self interstitials) under irradiation often interact with the solute elements present in the materials. Solutes can precipitate and/or segregate on point defect clusters (loops or voids) or extended defects (dislocations, grain boundaries). These modifications of the microstructure affect directly the mechanical properties of the materials. Thus, modelling should take into account the most important solute elements in the chemical composition of the industrial alloys. For the pressure vessel steels (for which an important international efforts is done in particular thanks to the PERFECT [2] and PERFORM-60 european projects) the evolution of the microstructure of dilute Fe alloys as complex as Fe-CuNiMnSiP-C under irradiation are modelled using a multiscale approach based on ab initio, molecular dynamics and kinetic Monte Carlo (KMC) simulations. In these atomic KMC simulations, both self interstitials and vacancies, isolated or in clusters, as well as carbon atoms are modelled [3]. The short term evolution of the microstructure is simulated. The medium to long term evolution of the microstructure is obtained by object KMC and cluster dynamics, considering a “grey” material. The interaction of some of these defects with dislocations are characterised by molecular dynamics in order to be used in mesoscopic dislocation dynamics. A

similar approach is developed for austenitic materials modelled by a concentrated FeCrNi alloy (γ-

Fe70Cr20Ni10). The thermal ageing (without irradiation) of FeCr alloys will also be presented. In the framework of the european projects dedicated to the pressure vessel steels and the austenitic steels, the multiscale modelling methods of the microstructure have been capitalised within two tools (RPV and INTERN) [4]. References [1] C. S. Becquart, C. Domain, Metallurgical and Materials Transactions A 42 (2011) 852. [2] Spetial Issue: PERFECT project. Journal of Nuclear Materials, 406 (2010). [3] C.S. Becquart, C. Domain, Phys. Status Solidi B 247 (2010) 9. [4] G. Adjanor et al., J. Nucl. Mater 406 (2010) 175.


37

Cluster Dynamics Modeling of Cavity and Loop Microstructure under Irradiation

Thomas JOURDAN1

1CEA-DEN-DMN, Service de Recherches de Métallurgie Physique, SRMP (Saclay, France)

The simulation of the loop and cavity microstructure evolution over long time scales requires efficient coarse grained methods. Cluster dynamics belongs to this class of methods, inasmuch as doses of around 100 dpa can be simulated in reasonable computation times. When switching to coarser grain scales, one often needs to resort to some approximations, either in the physical model or in the modeling techniques. Obviously these approximations must be controlled as carefully as possible in order to preserve the same degree of reliability from one scale to the other. In the case of cluster dynamics, the method operating at finer scale is object kinetic Monte Carlo (OKMC), which considers the microstructure as a collection of objects interacting with each other. The physical model is the same as cluster dynamics but the modeling technique is very different: in the case of OKMC, clusters are explicitly considered, whereas in cluster dynamics, only cluster concentrations are retained. Under thermal aging, as long as clusters can be considered as uniformly distributed in a representative volume, OKMC and cluster dynamics are shown to be rigorously equivalent. Under ion or neutron irradiation, displacement cascades are generated, which results in significant spatial correlations between defects. The simulation of such irradiation conditions is straightforward in OKMC, but is often seen as the crux of the problem in cluster dynamics. It is all the more important as physical results are often very sensitive to the irradiation term, since it can control entirely the cluster density through the nucleation of clusters inside cascades. To solve this problem, we have recently proposed a method to introduce cascades in cluster dynamics without loss of information compared to OKMC [1]. This method is based on a homogenization technique which makes use of a modified version of OKMC (Figure 1). The modified OKMC provides an effective source term for cluster dynamics, which is shown to reproduce satisfactorily reference calculations performed with full OKMC calculations.


38

Figure 1: Snapshot of a modified OKMC simulation used to homogenize a cascade.

The defects are the small colored spheres. References [1] T. Jourdan and J.-P. Crocombette, “Rate theory cluster dynamics simulations including spatial

correlations within displacement cascades”, Phys. Rev. B 86 (2012) 054113.


39

Phase and Microstructure Evolutions under Irradiation: Design of Coarsening-resistant Nanostructures

Pascal BELLON1, Robert S. AVERBACK1

1Department of Materials Science and Engineering, University of Illinois at Urbana-Champaign (Urbana, IL, USA)

Advanced nuclear energy systems will require materials that resist to very large doses of radiation damage. One strategy to accomplish that goal is to employ nanostructured materials, as their high density of sinks can dramatically enhance the trapping and the recombination of point defects, thus minimizing deleterious effects such as radiation-induced segregation and precipitation or swelling. A concern however is that these nanostructures would not be stable under irradiation and would undergo significant coarsening. We will discuss two possible approaches for designing coarsening resistant nanostructures. The first approach relies on the spontaneous formation of nanostructures at steady-state under irradiation in immiscible alloys, when the chemical mixing forced by nuclear collisions compete with the decomposition promoted by thermally activated decomposition [1]. These nanostructures are thus, by design, stable under irradiation. Examples of such nanostructures are found in ion-irradiated Cu-Ag, Cu-Fe and Cu-Co alloys [2], as seen in Fig. 1. A second approach relies on the formation of nanoscale precipitates leading to a microstructure that is kinetically trapped, owing to very low solubility and diffusivity of the precipitating elements in the matrix. We successfully applied this approach by alloying Cu with refractory solute elements such as Nb, Mo, and W, and show that nanostructures in these systems can resist coarsening at temperatures up to 85% of the melting point of Cu [3], as illustrated in Fig. 2. We show that this approach offers rich possibilities by considering ternary alloy systems [4], and that, in addition to thermal annealing, irradiation or severe plastic deformation can be used to synthesize nanostructures optimized for coarsening resistance at high temperature or under irradiation.


40

Fig. 1: Irradiation-induced nanopatterning in Cu90Ag10 thin film irradiated with 1.8 MeV

Kr ions at variable temperatures [2].

Fig. 2: Z-contrast TEM image of nanostructured Cu-Nb-W alloy with high

resistance to thermal coarsening at 650˚C.


41

References

[1] R. Enrique, P. Bellon. Phys. Rev. Lett. 84, 2885 (2000). [2] S. W. Chee, B. Stumphy, N. Q. Vo, R. S. Averback, P. Bellon. Acta. Mater. 58, 4088-4099 (2010). [3] N. Q. Vo, S. W. Chee, D. Schwen, X. Zhang, P. Bellon, R. S. Averback. Scripta. Mater. 63, 929-932

(2010).

[4] X. Zhang, N. Q. Vo, P. Bellon, R. S. Averback. Acta. Mater. 59, 5332-5341 (2010).


42

Session 3

STRUCTURAL, MECHANICAL, CHEMICAL CHARACTERIZATION


43


44

Nuclear Reactor Fuels: Materials with Highly Complex Behaviour

Rudy J. M. KONINGS1, 2, Dario MANARA1, Ondrej BENEŠ1, Christine GUÉNEAU3

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

2 Delft University of Technology, Faculty of Applied Sciences (Delft, The Netherlands)


This lecture will principally focus on the high temperature characterisation of nuclear fuel materials, which is a challenging task. Nuclear fuels are highly complex materials. They operate under extreme conditions such as high temperatures (up to 1500 K for conventional Light Water Reactors, 2400 K for Fast Neutron Reactors), in intense radiation fields and undergo significant changes in chemical composition during their life. In case of accidental conditions, the fuel may experience even more extreme conditions that may result in melting or reactions with cladding and/or coolant. This will be explained in a general introduction.

. Fig. 1: The UO2-PuO2 section of the ternary U-Pu-O phase diagram optimized by CALPHAD [1] taking into account recent experimental results. The UO2-PuO2 solid solution is indicated as

MOX. Empty and full red circles represent new experimental results [2]. The main part of the lecture will address the high temperature properties of UO2 and PuO2, the main components of current nuclear fuels, and their solid solution. Also the high temperature properties of the minor actinide oxides will be discussed, in view of their recycling in fast reactor fuels (transmutation). Extensive studies have been performed on particularly UO2 and PuO2 in 1950s and 1960s that have laid the foundation for our current understanding of these materials. The available data from that research has been gathered in numerous assessments but are not always conclusive. This is generally due to the high oxygen potential and the reactivity with container materials that might affect experiments at high temperatures. For that reason we have performed new experimental studies, when possible using innovative experimental approaches.


45

The experimental work includes melting point studies by laser heating (self-crucible), vapour pressure measurements by Knudsen effussion mass spectrometry, and calorimetry, and yielded is some cases remarkable differences with the literature [2]. For the interpretation of the results a close link with thermodynamic modelling has proven to be extremely useful, as will be demonstrated. As an example the optimised phase diagram of the UO2-PuO2 system, based on new results from laser heating experiments and the CALPHAD assessment, is given in Figure 1. References [1] C. Guéneau, N. Dupin, B. Sundman, C. Martial, J.-C. Dumas, S. Gossé, S. Chatain, F. De

Bruycker, D. Manara, R. J.M. Konings. J Nucl Mater 2011;419: 145-165. [2] F. De Bruycker, K. Boboridis, R.J.M. Konings, M. Rini, R. Eloirdi, C. Guéneau et al.. J Nucl Mater

2011; 419: 186-194.


46

Post-Irradiation Analysis of Fission Gases in Nuclear Fuels

Christophe VALOT1, Jean NOIROT1, Yves PONTILLON1

1CEA-DEN-DEC, Service d’Analyse et de Caractérisation du Comportement des Combustibles, SA3C (Cadarache, France)

Researches about irradiated fuel are conducted in order to investigate fuel properties and fuel behavior. These works cover various nuclear fuel operating conditions (i.e. Nominal operating conditions, off normal and severe accidents conditions, storage conditions…). Experimental investigations may be carried out through hot cell characterization and/or analytical tests [1-3], (Fig. 1). This talk is focused on the Fission Gas (FG) behavior which is a leading parameter in the overall fuel rod performance. Several parameters promote fission gas release (FGR) from nuclear fuels. Temperature is the most important one, but the local fuel microstructure has also to be considered as a key parameter. With the increase in fuel burn-up, the formation of a FG rich microstructure so-called High Burn-up Structure (HBS) has been evidenced in the fuel pellet periphery. Thus, it is a key issue to assess the contribution of the different parts of the fuel pellet to the total FGR using Post-Irradiation analyses. Through an example of experimental program performed on a standard PWR UO2 fuel, the benefits of a coupled approach "Post-Irradiation analyses" and "out-of-pile annealing test" is shown.

(a)

(b) (c)

Fig. 1: LECA-STAR Hot-cell capabilities; (a) µ-analysis laboratory; (b) puncturing facility; (c) Annealing tests facility (MERARG II).


47

References [1] J. Noirot et al., LWR fuel gas characterization at CEA Cadarache LECA-STAR Hot Laboratory.

Post-Irradiation Examination and In-Pile Measurement Techniques for Water Reactor Fuels, IAEA-TECDOC-CD-1635-2009.

[2] J. Noirot et al., High burnup changes in UO2 fuels irradiated up to 83 GWd/t in M5® claddings. Journal of Nuclear Engineering and Technology, Vol. 41, n°2-3 (2009).

[3] Y. Pontillon et al., Experimental and Theoretical Investigation of Fission Gas Release from UO2 up to 70 GWd/t under Simulated LOCA type Conditions: the GASPARD Program. Proceeding of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22, 2004.


48

Impact of Fuel Assembly Transportation on Zirconium Alloys: toward a Mechanistic Understanding

Fabien ONIMUS1, Joël RIBIS1, Brice BOURDILIAU2, Chantal CAPPELAERE2

1 CEA-DEN-DMN, Service de Recherches Métallurgiques Appliquées, SRMA (Saclay, France)

2 CEA-DEN-DMN, Service d’Etudes des Matériaux Irradiés, SEMI (Saclay, France)

Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are transported for wet storage to a devoted site. During dry transportation, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. A recovery of the radiation damage can occur during transportation that can affect the subsequent mechanical properties [1]. The recovery of the radiation damage during heat treatments has been investigated using micro-hardness tests at room temperature on neutron irradiated cladding materials made of fully recrystallized Zr-1%Nb alloy. Transmission electron microscopy (TEM) observations performed on irradiated thin foils have also shown that, simultaneously with the recovery of the hardness, the dislocation loop density, induced by irradiation, falls while the loop size increases (Fig. 1). Moreover, the TEM analysis has revealed that only vacancy loops are present in the material after long-term annealing, the interstitial loops having entirely disappeared. A numerical cluster dynamic modeling [2] has been used in order to reproduce the material recovery for various annealing conditions (Fig. 1).

Figure 1: Evolution of : (a) the loop density and (b) the loop diameter during a thermal annealing at 350°C, 400°C and 450°C. Experiment (Exp) and modeling (Sim).

Furthermore, the mechanical behavior of the cladding after post-irradiation creep test has been investigated. Creep tests under internal pressure were conducted at 400 and 420°C on the neutron irradiated recrystallized Zr-1%Nb alloy. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing [3]. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material (Fig. 2).


49

The transmission electron microscopy examinations, conducted on ring samples, revealed that the radiation defects have been annealed (Fig. 3). It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material [4]. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation.

0

100

200

300

400

500

600

700

800

0% 10% 20% 30% 40% 50%

Strain (%)

Str

ess (

MP

a)

ZrNb-4 (as-irradiated)

ZrNb-2 (400 °C - 130 MPa - 984 h)

ZrNb-1 (400 °C - 130 MPa - 3,301 h)

ZrNb-3 (420 °C - 130 MPa - 912 h)

ZrNb-5 (non irradiated)

Figure 2: Stress-strain curve obtained during ring tensile test on Zr-1%Nb alloy at room

temperature.

Figure 3: Dislocation microstructure after ring tensile test subsequent to a post-

irradiation creep test at 420°C. References [1] C. Ferry, C. Poinssot, C. Cappelaere, L. Desgranges, C. Jegou, F. Miserque, J.P. Piron, D.

Roudil, J.M. Gras. Journal of Nuclear Materials, 2006, Volume: 352, Pages: 246–253. [2] J. Ribis, F. Onimus, J.-L. Béchade, S. Doriot, A. Barbu, C. Cappelaere, C. Lemaignan. Journal of

Nuclear Materials, 2010, Volume: 403, Pages: 135–146. [3] B. Bourdiliau, F. Onimus, C. Cappelaere, V. Pivetaud, P. Bouffioux, V. Chabretou, A. Miquet.

Journal of ASTM International, 2011, Vol. 7, No. 9 (16th International Symposium on Zirconium in the Nuclear Industry, ASTM STP 1529, M. Limbäck, P. Barberis Eds, West Conshohocken, PA, 2011, pp. 929-953).

[4] F. Onimus, I. Monnet, J.-L. Bechade, et al. Journal of Nuclear Materials, 2004, Volume: 328, Issue: 2-3, Pages: 165-179.


50

Nuclear Materials Characterization by Tomographic Atom Probe

Philippe PAREIGE1, Bertrand RADIGUET1, Cristelle PAREIGE1

1CNRS, Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen

Faculté des Sciences Université de Rouen (Saint Etienne du Rouvray, France)

Nuclear Materials (reactor vessel, internal structures, fuel rod, glass containment…) undergo degradations under neutron irradiation due to particle/matter interaction. These degradations modify their physical, chemical and mechanical properties and are likely to impact the safety or the operation length of the concerned structures. It is thus crucial to anticipate and quantify these effects in order to ensure an efficient use of present or future nuclear reactors. The phenomena controlling materials behavior under irradiation are on an atomic scale. Understanding the phenomena on this scale is thus essential for the development of a predictive simulation and understanding of the ageing and safety of materials. This atomic scale is the first floor of the multi scale approach necessary for modeling and prediction, as illustrated in figure 1. The major technique in this field is the Tomographic Atom Probe, nanoanalytical technique with atomic resolution. Atom Probe Tomography (APT) is an extension in 3D of the atom-probe field ion microscope (APFIM), an instrument designed in the late sixties by E.W. Müller [1]. APT is the only approach able to map out the 3D distribution of chemical species in a material at the atomic-scale. The principle of APT [2,3] is based on the field evaporation of surface atoms of the specimen (a sharply pointed needle, R ~ 50 nm) and the chemical identification of field evaporated ions by time-of-flight mass spectrometry. The position of atoms at the sample surface is derived from the impact position of ions striking the detector. A major advantage of APT is its quantitativity. The local composition in a small selected region of the analyzed volume is simply derived from the number of atoms of each observed species. This is a big advantage compared to number of other instruments. No calibration is required. The in-depth resolution, independent of technologic details of these instruments, reaches 10 picometers. Atomic planes can therefore be imaged and the chemical order can be both exhibited and characterised. The lateral resolution is close to 0.5 nm. This makes it possible the quantitative analysis of small precipitates in alloys or the study of solute segregation to crystal defects (as illustrated in figure 2 [4]). The last few years have been the witness of a major breakthrough in the development of APT. Formerly limited to metals or good conductors, the implementation of ultra-fast pul-sed laser to the instrument opened APT to semi-conductors or oxides [5, 6]. Laser pulses give rise to very rapid thermal pulses that promote the field evaporation of surface

atoms.

Space

1 nm 1 m

Time

1 s

1 ps40 nm

Atoms

Electrons

Microstructure

10 mm

Dislocations andirradiation defects

Structure

1mm

1 y

1m

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10 m

Electronic StructureMolecular DynamicsObject or Event Monte Carlo

Crystal Plasticity (CP)Homogenization

Finite Elements (EF)

Monte CarloClusters Dynamics

Formation and mobilityof point defects (dp)

Evolution of Dislocations Networkdefects clusters, solute

Multi-scale Modeling

Dislocations Dynamics (DD)

TEM

TAP

SANS

MechanicalTestsSEM

EBSD

Déformation

Behavior rules

Ab initio

Fig. 1: Development of numerical tools and experiments for a multiscale approach.

Fig. 2: Tomographic Atom Probe experiment in a RPV steel. Each dot represents an atom (Si or P here). Irradiation indu-ced clusters and segregation on dislocation lines are visible.


51

In addition, a wider field of view is achieved so that larger area of analysis are obtained (~100x100 nm2) improving therefore statistics together with shortening the number of analyses required to get the relevant information. This technique brings key information for the understanding of the evolution of the microstructure of nuclear reactor steels under neutron irradiation, evolution responsible for their hardening and embrittlement. The basic principles of the technique, the sample preparation and some examples of applications in the field of “Metals under Irradiation” [7,8,9] will be presented and discussed.

References [1] E.W. Müller, J. Panitz, And S.B. Mc Lane. Rev. Scie. Instrm.: 39, 83 (1968). [2] A. Cerezo, I. J. Godfrey And G. D. W Smith. Rev. Sci. Instrum: 59, 862-866 (1988). [3] D. Blavette, A. Bostel, J.M. Sarrau, B. Deconihout And A. Menand: Nature, 363, 432-435 (1993). [4] Huang Hefei, Influence de l’irradiation aux neutrons sur le vieillissement des aciers de cuves des

réacteurs nucléaires à eau pressurisée. PhD 2012, Rouen University. [5] B. Deconihout, F. Vurpillot, B. Gault, G. Da Costa, M. Bouet, A. Bostel, D. Blavette, A. Hideur, G.

Martel: M. Brunel. Surf. Interface Anal. 39, 278, (2007). [6] B. Gault, F. Vurpillot, A. Vella, M. Gilbert, A. Menand, D. Blavette. B Rev. Sci. Instr. 77, 043705

(2006). [7] A. Etienne, B. Radiguet, N.J. Cunningham, G.R. Odette, R. Valiev, P.Pareige, Comparison of

radiation induced segregation in ultrafine-grained and conventionnal 316 austenitic stainless steels. Ultramicroscopy 111 (2011) 659-663.

[8] M. Brocq, B. Radiguet, S. Poissonnet, F. Cuvilly, P. Pareige, F. Legendre, Nanoscale characterization and formation mechanism of nanoclusters in an ODS steel elaborated by reactive-inspired ball-milling and annealing. Journal of Nuclear Materials 409 (2011) 80-85 (doi:10.1016/j.jnucmat.2010.08.043).

[9] 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, Vol.415 Issue: 1 Pages: 61-66, 2011.


52

Mechanical and Thermal Resistance of Multi-Material Components for ITER

Hélène BURLET1

1CEA-DEN-DRT, Laboratoires d’Innovation pour les Technologies des Energies, LITEN (Grenoble, France)

The First Wall panels for ITER are complex parts composed of stainless steel, copper and beryllium [1]. These materials are joined using diffusion bonding technique. The stainless steel is a commonly used in nuclear reactors 316LN material and acts as a structural material. The copper alloy is a CuCrZr material which acts as a heat sink. The beryllium consisting in tiles and layer is used as the protective plasma facing material. The fabrication of these panels is performed through 2 main steps. The first step consists in welding all together a bi-metallic support structure made from a thick CuCrZr plate embedded with 316LN tubes and bonded to a thick 316LN backing plate with cooling channels. The bonding is performed in a HIP (Hot Isostatic Pressure) facility. The second step is performed at a lower temperature and aims at simultaneously welding by HIP Be onto CuCrZr and ageing the CuCrZr heat sink to obtain the correct mechanical resistance of this alloy reinforced by precipitates. The various joints 316LN/316LN, 316LN/CuCrZr, and CuCrZr/Be are then characterized [2] from a microstructural point of view and by mechanical tests. It is quite hard to characterize the strength of a diffusion bonded joints. Standard tests may be used for homogeneous joints whereas specific tests have been developed to characterize the heterogeneous bonds. To optimize the bond, we performed mainly impact and tensile bimaterial tests (Fig 1). Once the manufacture parameters have been optimized, advanced mechanical tests are performed based on Bimetallic CT specimens, axisymmetric notched specimens, 4P bending specimens. Numerical simulations are required to analyse the mechanical response. In order to characterize the fatigue resistance of the joints, specific mock-ups have been designed by the European Fusion Development Agreement – EFDA – team (Fig 2). Results of heat flux testing will be reviewed for the various joints. The assembly of heterogeneous materials by HIPping is very complex. It may be used for various applications appart ITER FW panels such as test blanket module and divertor. It is also envisaged in PWR fission reactors, for instance to join 316 stainless steel onto 16MND5.

Fig. 1: Tensile tests specimens CuCrZr/316LN. Fig. 2 : Mock-ups for high heat flux tests.


53

References [1] Boudot, C., Boireau, B., Cottin, A., Lorenzetto, P., Bucci, P., Gillia, O., Manufacture of a shield

prototype for primary wall modules. Fusion Engineering and design, Volume 83, Issues 7-9, December 2008, pp. 1294-1299.

[2] Gillia, O., Briottet, L., Chu, I., Lemoine, P., Rigal, E., Peacock, A., Characterization of CuCrZr and CuCrZr/SS joint strength for different blanket components manufacturing conditions. Journal of Nuclear Materials, Volumes 386-388, 30 April 2009, pp. 830-833.


54

New Characterizations at the MARS Beamline (SOLEIL Synchrotron Radiation)

Jean-Luc BÉCHADE1, Denis MENUT1

, Bruno SITAUD2, Sandrine SCHLUTIG2, Isabelle LLORENS2, Marie-Laure LESCOAT1, Joel RIBIS1, Nicolas JONQUÈRES3,

Dominique LETERME4,


2Synchrotron SOLEIL, Ligne de MARS (Saint Aubin, France)

3CEA-DEN-DM2S, Service d’Etudes Mécaniques et Thermiques, SEMT (Saclay, France)

4CEA-DEN-DANS, Service de Soutien aux Projets, à la Sécurité et à la Sûreté, SP2S (Saclay, France)

MARS (Multi-Analyses on Radioactive Samples) is the X-ray bending magnet beamline of the French synchrotron facility SOLEIL dedicated to the study of radioactive matter by means of a multi-technique equipment [1]. The MARS beamline is the fourth dedicated beamline for studying radionuclides in Europe (after the ROBL beamline at ESRF (Grenoble, France), the INE beamline at ANKA (Karlsruhe, Germany) and the microXAS beamline at SLS (Villingen, Switzerland) and aims at extending the possibilities of

synchrotron based X-ray characterizations towards a wider variety of radioactive elements ( and n emitters). Thus, its specific and innovative infrastructure has been optimized to carry out analyses on radioactive materials with global activities up to 18.5 GBq per sample [2]. Particularly, it

offers unique possibilities for studying and neutrons emitters’ samples with an activity up to 2 GBq. Nevertheless, such a possibility supposes to obtain special authorizations from French safety authorities (ASN) based on specific technical innovations to assure the confinement of samples and to reduce the dose rate during their manipulation on the beamline. This beamline, which has been built thanks to a close partnership and support by the CEA, has been designed to provide X-rays in the energy range of 3.5 keV to 35 keV. Three main techniques are progressively proposed on MARS beamline: transmission and high-resolution powder X-ray diffraction (respectively TXRD and HRXRD), X-ray absorption spectroscopy (XAS) and X-ray fluorescence (XRF) [3]. After the preliminary experiences performed in 2009-2010 on un-irradiated samples [1, 3, 4], this presentation deals with recent results obtained at the MARS beamline, thanks, (i) to very powerful and useful improvements brought to the experimental set-up of the beamline (especially High Resolution diffractometer, fig. 1) and (ii) to various materials irradiated at high doses with ions (ODS up to 200 dpa) and also Zr based alloys irradiated with neutrons in Pressurized Water Reactors up to 5 PWR cycles. The results obtained by X ray Diffraction on secondary phases evolutions as a function of irradiation doses for both ODS and Zr based alloys will be exposed, along with the very first XAS experiences performed on ODS materials. Finally, future prospects and main objectives concerning the evolution of the beamline for our program on irradiated materials will be discussed; for example the milestone for 2013 concerning the analysis on ODS irradiated at high doses with neutrons.


55

MAR 345

Imaging plate

Multi-crystalanalyzer with 24

channelscomposed of

Ge(111) crystals Diffractometerwith twocoaxial high precision rotation

stages( , 2 )

Goniometer with :

-Two high precision ( , ) rotations (sphere of confusion of 40 µm in diameter)

-- Three translation stages (Tx, Ty, et Tz)

sample

Incoming monochromatic X-ray beam

Fig. 1: High Resolution X Ray Diffractometer at the MARS beamline.

References [1] B. Sitaud, P.L. Solari, S. Schlutig, I. Llorens, H. Hermange, Characterization, of radioactive

materials using the MARS beamline at the synchrotron SOLEIL. Journal of Nuclear Materials 425 (2012) 238-243.

[2] P.L. Solari, S. Schlutig, H. Hermange and B. Sitaud. Journal of Physics: Series 190 (2009) 012042.

[3] S. Schlutig, P.L. Solari, H. Hermange, B. Sitaud, Basic Actinide Science and Materials for Nuclear Applications, 2010. Materials Research Society Symposium Proceedings, 1264, p.131-136.

[4] J.-L. Béchade, D. Menut, M.-L. Lescoat, B. Sitaud, S. Schlutig, P.L. Solari, I. Llorens, H. Hermange, Y. de Carlan, J. Ribis, L. Toualbi, Application of synchrotron radiation to analyze the precipitation in ODS materials before irradiation in Fe–9%Cr single grain of powder and consolidated Fe–18%Cr. Journal of Nuclear Materials 428 (2012) 183-191.


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Session 4

NUCLEAR MATERIALS AGEING


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58

Materials Ageing Degradation Programme in Japan and Proactive Ageing Management in NPP

Tetsuo SHOJI1

1Tohoku University, New Industry Creation Hatchery center, LMRR unit (Sendai, Japan)

Predictive and preventive maintenance technologies are increasingly of importance for the long term operation (LTO) of Light Water Reactor (LWR) plants. In order for the realization LTO to be successful, it is essential that aging degradation phenomena should be properly managed by using adequate maintenance programs based on foreseeing the aging phenomena and evaluating their rates of development, where Nuclear Power Plants can be continued to operate beyond the original design life depending upon the regulatory authority rules. In combination with Periodic Safety Review (PSR) and adequate maitenance program, a plant life can be extended to 60 years or more. Plant Life Management (PLiM) is based upon various maintenance program as well as systematic safety review updated based upon the state of the art of science and technology. One of the potential life time limiting issue would be materials ageing degradation and therefore an extensive efforts have been paid worldwidely. In 2007, NISA launched a national program on Enhanced Ageing Management Program and 4 nationwide clusters were formed to carry out the national program where materials ageing degradation was one of the major topics. In addition to these degradation modes, one important activities in this program is proactive materials degradation management directed by the autor which is a kind of the extension program of NRC PMDA program based upon more fundamental approach by a systematic elicitation by the experts nominated from all over the world. NISA program can be devived into two phases, one is from fiscal years (FY) 2006 - 2010 and the other FY 2011. Later phase is focusing more on System Safety due to Fukushima NPP accident. The main objectives of the Phase I is to evaluate potential and complex degradation phenomena and their mechanisms in order to identify future risks of component aging in nuclear power plants. The following items are of particular concern in this phase: (a) investigation of potential materials ageing phenomena and corresponding plant issues, and (b) investigation of the effectiveness of evaluation techniques, concerning potential aging phenomena. In NISA PMDM Phase I, three approaches are considered to be essential for proactive aging management. First one is a deductive science- based approach, second one is an intuitive based approach based upon a careful analysis of operating experiences and third one is systematic elucidation. In particularly, the deductive method based on a fundamental scientific understanding of material degradation is highlight for the management of possible latent or cascade materials degradation processes which have not yet become obvious in operating plants. Proactive management issues associated with materials aging in LWRs were discussed in terms of suggested research topics that should be undertaken in either the short or long-term. Based on these discussions prioritized lists of medium and long term research projects were established for both PWRs and BWRs. In this paper, the research subjects to be considered for the aging degradation phenomena in PWR structural materials, which were discussed at the Proactive Aging Management Experts' Panel Meeting of the NISA Project, are introduced as follows. 1. Pressurized Water Reactor (PWR) 1.1 Research subjects for shorter term projects to be completed in ~5 years

1) SCC initiation phenomena, including the effects of surface stress/strain, residual stress, microstructure and strain localization, 2) Development of qualified mechanisms-based lifetime models for PWSCC propagation in nickel base alloys, 3) Characterization of weld metals, dilution at interfaces and heat affected zones,


59

4) Strain localization, strain history and relationship to cold work, microstructure and compositional banding, 5) Effect of environment on fatigue and fracture resistance in PWRs, 6) Flow-accelerated corrosion in PWRs.

1.2 Research subjects for longer term projects to be completed in ~5 to 10 years. 1) Effects of irradiation flux and fluence on stainless steels and nickel base alloys and the effect on SCC, 2) Modeling and validation of residual stress/strain profiles in complex welded geometries and how these may change with neutron fluence, 3) Initiation of SCC in structural alloys, modeling stochastic features, heat to heat variability, effects of long exposure periods, 4) Quantification of potential synergistic effects between competing degradation modes for PWRs, 5) Quantitative modeling of oxidation and EAC based on fundamental physical principles for PWRs. In particular, oxidation localization and acceleration is a key process of these environmentally assisted cracking and we proposed a novel approach to characterize this oxidation localization behavior based upon the profile analysis of metal/oxide interface. Oxidation localization dynamics can be quantitatively evaluated by the parameter, depth of localized oxidation penetration. Analyzed results can be summarized as Fig. 1.

Fig. 1: Oxidation localization behavior and influence of surface finish (polished and lathed) and neutron irradiation on 316L stainless steel in oxygenated water.

References

[1] J. Muscara, Expert Panel Report on Proactive Materials Degradation Assessment. USNRC NUREG Report, CR-6923, February 2007. [2] T. Shoji et al., Proactive Material Degradation Research Subjects for Light Water Reactors. E-Journal of Advanced Maintenance Vol. 4, No. 2 (2012) pp. 36 - 56, Japan Society of Maintenology.

Acknowledgements

This work has been performed as part of the Program of Enhancement of Ageing Management and Maintenance of Nuclear Power Plants organized by the Nuclear and Industrial Safety Agency, NISA, in the Ministry of Economy, Trade and Industry, METI, of Japan. The authors would like to thank the Experts' Panel members for their contributions without which no success of the program would never be made.

Tohoku University

Oxidation localization parameters

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・Oxidation localization develops with time especially in the specimens under

irradiation and lathe finished surface conditions (increase the parameter values)

・More significant oxidation localization is pointed out in the specimens under

irradiation

16

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Polished Lathed

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60

The Irradiation-Assisted Stress Corrosion Cracking (IASCC) Issue: some Examples of Studies Carried out at CEA

Benoît TANGUY1, Maxime SAUZAY2, Christian ROBERTSON2, Stéphane PERRIN3

1CEA-DEN-DMN, Service d’Etudes des Matériaux Irradiés, SEMI (Saclay, France)



Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors (radiation induced segregation (RIS) at grain boundaries, radiation induced microstructure (dislocations loops, voids, bubbles, phases), localized deformation under loading, irradiation creep and transmutations) evolutions due to neutron irradiation. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Fundamental studies related to the IASCC mechanisms can be divided on two main parts: (i) How the irradiation modifies the austenitic stainless steels (ASS) microstructure (and so the ASS mechanical behavior) as a function of dose, temperature, stress, spectrum and flux and how it affects the resistance of the ASS to SCC sensitivity and (ii) How the irradiation flux modifies the corrosion process themselves (oxidation) and how it may affect the SCC kinetics. For this last topic, both a modification of water chemistry (radiolysis) and a change of oxide are probably involved. To answer these questions, parallel to experimental characterizations at different scales, development of modeling is of growing importance in the understanding of the basic mechanisms of IASCC but also of their interactions. This lecture describes some studies carried out at CEA in order to provide further understanding in the IASCC damage modeling. First part of this lecture describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop [1] dedicated to SCC tests on neutron irradiated materials is then given. The main steps of the interrupted SCC tests carried are described relative to their objectives. Second part of the lecture gives some insights of the effect of irradiation on the oxidation processes. Oxidation of stainless steel in PWR primary water at 325°C has been studied by investigating the influence of defects created at the alloy subsurface by proton irradiation performed before corrosion test. Corrosion experiments were performed during two different corrosion sequences using H2

18O for the second corrosion one. The oxide layer was studied by SEM and TEM and could be divided in two parts: an external discontinuous layer composed of crystallites rich in iron and an internal continuous one richer in chromium. Tracer experiments underlined that the growth of this protective scale was due to oxygen diffusion in the grain boundary of the oxide layer. Defects created by irradiation have an effect on the two oxide layers. They are preferential nucleation site for the external layer and so increase the density of the crystallites. They also induce a slower diffusion of oxygen in the internal layer.

The third part of the lecture focuses on the plasticity of the irradiated stainless steel at the grain scale. The goal of this work is to model irradiation-induced strain localization at the grain scale, using 3D dislocation dynamics (DD) simulations. More specifically, it is attempted to predict the number of shear bands affecting (deforming) the grain boundaries, in presence of a representative irradiation defect cluster populations.


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In practice 2 types of DD simulations were used, based on their complementary capacities and limita-tions. Simulations (Figure 1) where irradiation-induced defect clusters are treated as planar obstacles to dislocation motion were carried out. This description has reduced computational load and is compatible with the introduction of thermally activated cross-slip for simulating multiple slip band formation. Shear band spacing and plastic strain spreading obtained using these simulations show that spacing increases with increasing dose, grain size and increasing stacking fault energy (SFE). The proposed model has been successfully extrapolated to grain sizes and defect cluster populations representative of actual fcc alloys, submitted to typical PWR neutron irradiation conditions.

Fig. 1: Clear bands and strain localization predicted using Type-2 DD simulations. Left image: the represented loop-facets are those potentially absorbed by mobile screw dislocations.

High density of absorbed loops materializes the channel position. Central image: swept loop-facets and corresponding dislocation structures. Right image: plastic strain mapping in grain

boundary corresponding to left and central images.

Finally, simulation results to assess the influence of slip localization effect, which is an important feature of plasticity in irradiated ASS’s, on the grain boundary stress-strain fields are presented. Slip localization is known to trigger grain boundary brittle fracture. For predicting the local stress fields, an elastoplastic slip band is assumed to be embedded at the surface of an elastic matrix. Numerous FEs computations have been carried out allowing the proposal of analytical formulae describing the grain boundary (GB) normal stress fields depending on clear band length and thickness, Schmid factor, slip band critical shear stress and remote stress mainly. Finally, finite fracture mechanics is applied, together with both critical GB fracture energy and stress criteria. This leads to analytical formulae as simple as the ones deduced from the pile-up theory, but taking into account the channel thickness. References [1] B. Tanguy, C. Pokor, A. Stern, P. Bossis, Initiation stress threshold Irradiation Assisted Stress

Corrosion Cracking criterion assessment for core internals in PWR environment. ASME Pressure Vessels and Piping Division Conference, July 17-21, 2011, Baltimore (MD), USA.

Acknowledgements

The studies presented in this paper have been performed within the frame of collaborative works with EDF and/or with the European program (FP7) PERFORM-60. The CEA DEN-RSTB, projects RACOC and MASOL are also acknowledged for financial and scientific support.


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Stress Corrosion Cracking of Nickel Base Alloys in PWR Primary Water

Catherine GUERRE1, Elizabeth CHAUMUN1, Jérôme CRÉPIN2, Ian de CURIÈRES3, Cécilie DUHAMEL2, Eva HÉRIPRÉ4, Emmanuel HERMS1, Pierre LAGHOUTARIS1,

Régine MOLINS2, Mohamed SENNOUR2, François VAILLANT5


2 MINES Paristech, Centre des matériaux, CNRS UMR 7633 (Evry, France)

3 Institute for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS (Fontenay-aux-Roses,France)

4 Laboratoire de Mécanique des Solides, Ecole Polytechnique (Palaiseau, France)

5EDF R&D – MMC (Moret-sur-Loing, France)

Stress corrosion cracking (SCC) of nickel base alloys and associated weld metals in primary water is one of the major concerns for pressurized water reactors (PWR). Since the 90’s, highly cold-worked stainless steels (non-sensitized) were also found to be susceptible to SCC in PWR primary water ([1], [2], [3]). In the context of the life extension of pressurized water reactors, laboratory studies are performed in order to evaluate the SCC behaviour of components made of nickel base alloys and of stainless steels. Some exemples of these laboratory studies performed at CEA will be given in the talk. This presentation deals with both initiation and propagation of stress corrosion cracks. The aims of these studies is, on one hand, to obtain more data regarding initiation time or crack growth rate and, one the other hand, to improve our knowledge of the SCC mechanisms. The aim of these approaches is to model SCC and to predict components life duration. Crack growth rate (CGR) tests on Alloy 82 with and without post weld heat treatment are performed in PWR primary water (Figure 1). The heat treatment seems to be highly beneficial by decreasing the CGR. This result could be explai-ned by the effect of thermal treatment on the grain boundary nanoscopic precipitation in Alloy 82 [4].

Fig. 1: Fracture surface of an Alloy 82 CT specimen tested 8064 hr at 325°C in simulated PWR primary water, view of an isolated intergranular SCC crack (Scanning Electron Microscopy).


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The susceptibility to SCC of cold worked austenitic stainless steels is also studied. It is shown that for a given cold-working procedure, SCC susceptibility increases with increasing cold-work ([2], [5]). Despite the fact that the SCC behaviour of Alloy 600 has been widely studied for many years, recent laboratory experiments and analysis ([6], [7], [8]) showed that oxygen diffusion is not a rate-limiting step in the SCC mechanism and that chromium diffusion in the bulk close the crack tip could be a key parameter. References [1] T. Couvant, P. Moulart, L. Legras, P. Bordes, J. Capelle, Y. Rouillon, T. Balon, PWSCC of

austenitic stainless steels of heaters of pressurizers, Proceedings of the Fontevraud V international symposium, Contribution of materials investigation to the resolution of problems encountered in pressurized water reactors, 23-27 september 2002.

[2] D. Feron, E. Herms, B. Tanguy, Behavior of stainless steels in pressurized water reactor primary circuits, Journal of Nuclear Materials, 2012, 427 (1-3) , pp. 364-377.

[3] C. Guerre, E. Herms, O. Raquet, S. Marie, M. Le Calvar, SCC crack growth rate of cold work austenitic stainless steels in PWR primary water conditions, 13th International Conference on Environmental Degradation in Nuclear Power Systems, Whistler, British Columbia, August 19 - 23, 2007

[4] C. Guerre, C. Duhamel, M. Sennour, J. Crépin, M. Le Calvar, SCC crack growth rate of Alloy 82 in PWR primary water conditions – effect of a thermal treatment, 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Colorado Springs, Colorado, August 7 – 11, 2011.

[5] E. Herms, O. Raquet, L. Séjourné, F. Vaillant, SCC of Cold-Worked Austenitic Stainless Steels Exposed to PWR Primary Water Conditions : Susceptibility to Initiation, EUROCORR 2009 – septembre 2009, Nice (France), paper 7861.

[6] M. Sennour, P. Laghoutaris, C. Guerre, R. Molins, Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment, Journal of Nuclear Materials 393 (2009) 254–266.

[7] C. Guerre, P. Laghoutaris, J. Chêne, L. Marchetti, R. Molins, C. Duhamel, M. Sennour, Stress corrosion cracking of Alloy 600 in PWR primary water : influence of chromium, hydrogen and oxygen diffusion, 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Colorado Springs, Colorado, August 7 – 11, 2011.

[8] P. Laghoutaris, C. Guerre, J. Chene, R. Molins, F. Vaillant, I. De Curieres, Contribution to model stress corrosion cracking of Alloy 600 in PWR primary water, Workshop on Detection, Avoidance, Mechanisms, Modeling, and Prediction of SCC Initiation in Water-Cooled Nuclear Reactor Plants, 07/09/2008 - 12/09/2008, Beaune, France.


64

Thermal Ageing Effects: Examples on Materials of PWR and Preventive Measures in the Design of EPRTM Plants

Pierre JOLY1, Pascal OULD1, François ROCH1

1AREVA NP – Engineering E&P, Division of Primary Components – Materials Department (Paris, la Défense, France)

Even though the operating temperature of Pressurized Water Reactors (PWR) is moderate (around 300°C), experience gained on some of the materials used for the manufacturing of components of existing PWRs, shows that they may be sensitive to thermal ageing phenomena. This type of ageing mechanism essentially causes an embrittlement, i.e. a reduction of fracture toughness, or an increase of the ductile to brittle transition temperature (for bainitic or martensitic steels), as the time in service increases. In the presentation, three examples of thermal ageing phenomena affecting PWR components are treated. The first one deals with the cast duplex austenitic-ferritic stainless steels, which are used in various locations of the main coolant piping such as elbows, nozzles, centrifugally cast straight sections, etc… (Fig. 1). The second example is about martensitic stainless steels, mainly used for bolting and internal components of valves (Fig. 2). Eventually, the third example is the case of bainitic low alloy steel, used to manufacture the large primary pressure vessels such as the Reactor Pressure Vessel, the Steam Generator, and the Pressurizer (Fig. 1). In all these examples, the presentation illustrates how these degradation mechanisms are known, through laboratory studies. These studies allow in particular deriving the main parameters of the material itself and of the service conditions which govern the embrittlement, and allow to derive predictive models of embrittlement, accurate enough for engineering applications (Fig. 3). In some cases this laboratory knowledge is complemented and validated by field experience, obtained through the surveillance of components in service, or through expertise programs conducted on decommissioned components. Finally, from the knowledge on these phenomena, accumulated in the research programs, examples of precautions taken in order to eradicate or to mitigate the known effects of thermal ageing in the design of new EPRTM plants are presented.

Fig. 1: Layout of the main components of the primary circuit of a 4 loop Pressurized Water Reactor.


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Fig. 2: Example of the effect of thermal ageing on the Charpy V Impact Energy transition curve

of a typical 17-4 PH martensitic stainless steel used for valve stems applications.

Fig. 3: Example of modelling of the kinetics of embrittlement of low alloy Pressure Vessel steel, at different temperatures by grain boundary segregation of Phosphorous.

Phosphorous Grain Boundary Coverage

0

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of embrittlement


66

Plasma Facing Components: Challenges For Nuclear Materials

Philippe MAGAUD1, Jérôme BUCALOSSI1, Antonella LI PUMA2, Marc MISSIRLIAN1, Marianne RICHOU1

1CEA-DSM-IRFM, Service Intégration Plasma-Paroi, SIPP (Cadarache, France)

2CEA-DEN-DM2S, Service d’Etudes des Réacteurs et de Mathématiques Appliqués, SERMA (Saclay, France)

In current fusion devices, the components located in front of plasma, the so-called plasma facing components (PFCs), sustain severe constraints such as high thermal flux (several MW/m²), erosion, flux of particles. The management of this first material interface is critical from a plasma performance point of view. ITER, as nuclear facility, is initiating a new era for fusion, which will be reinforced for a future fusion power plant which will add specific requirements (sufficient lifetime, a cooling system to produce energy, use of low activation material) while increasing nuclear constraints. The talk will recall in a first part the main requirements of an actively plasma facing components and the main results obtained with low-Z carbon based PFCs (mainly CFC). Experimental feedback from these challenging components is an essential step for the success of the next generation of components, in particular in term of manufacturing or handling intense heat loads. Nuclear safety requirements mainly drive the need of new materials for the nuclear phase of ITER. The tritium retention in carbon based PFCs and the strong erosion are expected to be too high in the Deuterium-Tritium phase with CFC targets, justifying the use of high-Z materials. The evolution toward high-Z materials, with tungsten the most promising, becomes a major challenge for fusion research. Large scale experiences with W have only been obtained recently with the operation of ASDEX-upgrade and JET tokamaks but with non-actively cooled PFCs. ASDEX-upgraded is equipped with W coated carbon PFCs while JET includes W coated carbon PFC and inertially cooled solid W, using in all cases a technology not relevant for ITER. Extensive R&D programmes have been performed in Europe to develop reliable actively PFCs for ITER [1-5]. The state of the art will be presented including specific devices needed to fully qualify, at laboratory scale, designs foreseen for ITER. In order to reduce the risks and anticipate any difficulties ITER may face in terms of manufacturing or operation, it is proposed to update Tore Supra with a full W first wall and divertor, benefitting from the unique long pulse capabilities of the Tore Supra platform, the high installed power and the long history of operation with actively cooled high heat flux components [6]. The main goals of the ‘WEST’ project (W - for tungsten -Environment in Steady-state Tokamak, figure 1) will be presented. The talk will also address acknowledged gaps in PFCs developments for DEMO, which require extensive studies in different topics, from plasma-surface interaction to engineering including material sciences. Further challenges address simultaneously the higher power density, high-temperature wall, bulk (neutron) and surface (charged particle) accumulated damage. The high neutron fluence expected in a fusion reactor (more than10 dpa/year) will affect erosion and tritium retention properties of materials. Near-surface material properties will be for instance altered by the neutron damage. Such synergistic effects are expected to be important in the DEMO environment and are difficult to be addressed experimentally. A more robust coupling of materials development, including fundamentally studies, with advanced design is required. If tungsten is the most promising material for the plasma-facing, tungsten also offers less favorable properties (recrystallization, which influences the mechanical properties, embrittlement as a result of neutron-induced damages, He-induced sputtering…) that have to be resolved. Only a global approach, including fundamental science, material development, joining/welding techniques, design innovation and a close link with plasma physics is able to reach the necessary level of credibility for operating such components in a fusion environment in an economically reasonable way.


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Fig. 1: The ‘WEST’ project is a major update of Tore Supra to address the issues related to tokamak operation with W actively cooled ITER relevant PFCs. Internal components, presently

in CFC will be updated to W components (ITER technology) and the magnetic plasma configuration (presently circular) will be changed to X-point configuration by installing a set

of poloidal coils inside the lower and upper parts of the vacuum vessel. References [1] - M. Missirlian, M. Richou, B. Riccardi, P. Gavila, T. Loarer and S. Constans, The Heat removal

capability of actively cooled plasma-facing components for ITER divertor. Physica Scripta, T145 (2011).

[2] - M. Richou, M. Missirlian, B. Riccardi,, P. Gavila, C. Desgranges, N. Vignal, V. Cantone and S. Constans, Fatigue lifetime of repaired high heat flux components for ITER divertor. Fusion Engineering and Design 86 (2011) pp. 1771-1775.

[3] - M. Missirlian, F. Escourbiac, A. Schmidt, B. Riccardi and I. Bobin-Vastra, Examination of high heat flux components for the ITER divertor after thermal fatigue testing. Journal of Nuclear Materials 417 (2011) pp. 597-601.

[4] - D. Serret, M. Richou, M. Missirlian and T. Loarer, Mechanical characterization of W-armoured plasma-facing components after thermal fatigue. Physica Scripta, T145 (2011).

[5] - A. Durocher et al., Infrared thermography inspection of the ITER vertical target qualification prototypes manufactured by European industry using SATIR. Fusion Engineering and Design 84 (2009) 314-318.

[6] – J. Bucalossi et al., Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing. Fusion Engineering and Design, Volume 86, Issues 6–8, October 2011, Pages 684-688.


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Irradiation Resistance in a Fusion Environment: a Challenge for Structural Materials

Jean HENRY1, Jean-Louis BOUTARD2

1CEA/DEN-DMN, Service de Recherche Métallurgiques Appliquées, SRMA (Saclay, France)

2CEA, Cabinet du Haut-Commissaire (Saclay, France)

The safe and reliable operation of future fusion power plants will require structural materials able to withstand highly demanding operating conditions. Indeed the materials of the First Wall, Tritium Breeding Blankets or Divertor, will be subjected to intense fluxes of high energy neutrons, up to high total doses and in a wide range of in-service temperatures. In this presentation, we will discuss the irradiation behavior of candidate steels for fusion application.

In the absence of an irradiation facility with a prototypical neutron flux, irradiation effects in these materials, such as the 9Cr tempered martensitic steel Eurofer, were up to now assessed based in particular on experiments using fission reactors. This approach has some relevance since both experimental data and Molecular Dynamics simulations indicate that the primary displacement damage in iron is very similar after irradiation by fission or 14 MeV neutrons. Furthermore damage rates in fusion devices will be very close to values typical of fast reactors such as BOR60. Like other 9Cr martensitic steels, Eurofer steel exhibits negligible swelling and very moderate hardening and embrittlement at irradiation temperatures above approximately 380-400°C. However results of irradiation experiments performed at 325°C up to about 80 dpa in BOR60 [1] showed an important increase of the yield stress in irradiated Eurofer and a significant degradation of the impact properties (large shift of the Ductile-to-Brittle Transition Temperature and decrease of the upper shelf energy). TEM and SANS investigations revealed that these modifications of the mechanical properties are due to the irradiation-induced formation of a high density of small dislocation loops and to alpha/alpha’ unmixing. In fusion irradiation conditions, significant amounts of helium and hydrogen will be produced in the materials in addition to displacement damage. Implantation experiments using a cyclotron have revealed that helium can induce a drastic embrittlement of 9Cr martensitic steels, with the occurrence of intergranular fracture [2, 3]. This fracture behavior is consistent with the results of electronic structure calculations which indicate that helium severely decreases grain boundary cohesion in iron. Likewise, measurements of tensile and impact properties after irradiation of 9Cr martensitic steels in a spallation environment show a degree of embrittlement significantly greater than that expected after irradiation in fission conditions, which is attributed in particular to the high helium content in the irradiated specimens [3]. Furthermore, results of dual (Fe+He) and triple (Fe+He+H) ion beam irradiations indicate that the high swelling resistance of Ferritic/Martensitic steels might not be maintained in a fusion irradiation environment. By contrast, Oxide Dispersion Strengthened (ODS) FM steels are promising materials for fusion application. In addition to lower irradiation-induced hardening at low temperature than FM steels [1], ODS displayed a better mechanical behavior after irradiation in spallation conditions [4]. Figure 1 shows an example of tensile curves measured on MA957 ODS irradiated in the SINQ spallation target. Irradiated MA957 retained significant ductility whereas FM steels subjected to identical irradiation conditions exhibited a brittle intergranular fracture mode. The high density of nanoclusters in ODS FM steels are believed to act as effective sinks for point defects as well as trapping sites for helium and hydrogen, which is confirmed by microstructural observations [5].


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0

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800

1000

1200

1400

1600

0 2 4 6 8 10 12 14 16 18 20

stre

ss (

MP

a)

strain (%)

Fig. 1: Engineering tensile curve measured at room temperature and SEM micrographs

showing the ductile fracture surface of MA957 ODS after irradiation at 360°C to 19 dpa in the SINQ spallation target. The accumulated He content in the irradiated specimen is about 0.17

at%. These data demonstrate the good mechanical behavior of MA957 following irradiation in an environment with high He/dpa ratio [4].

Finally it must be emphasized that the above mentioned data were not obtained in a fusion representative environment. Hence modeling of irradiation effects, which has made great progress in recent years as demonstrated by other presentations in this workshop, will be a necessary tool to extrapolate the experimental results to the real case and to optimize the experimental program to be conducted in a future intense 14 MeV neutrons source such as IFMIF. References [1] J. Henry, X. Averty, A. Alamo. J. Nucl. Mater. 417 (2011) 99. [2] R. Schäublin, J. Henry, Y. Dai. Comptes Rendus Physique, 9 (2008) 389. [3] Y. Dai, J. Henry, Z. Tong, X. Averty, J. Malaplate, B. Long. J. Nucl. Mater 415 (2011), 306. [4] J. Henry, X. Averty, Y. Dai, J.P. Pizzanelli, J.J. Espinas. J. Nucl. Mater 386–388 (2009) 345. [5] L. Hsiung, M. Fluss, S. Tumey, B. Choi, Y. Serruys, F. Willaime, A. Kimura. Phys. Rev. B 82

(2010) 184103.

Irradiated to 19 dpa, 0.17 at% He

As-received

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