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Home   /   E&T program   /   Continuous Professional Development   /   Metallurgy and properties of Zr alloys for nuclear applications

Metallurgy and properties of Zr alloys for nuclear applications

1.1.4 Matériaux pour le nucléaire

Materials

Materials for nuclear

Metallurgy

Nuclear fuel cycle

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In summary

The aim of this training is to present the main Zirconium-based alloys in pressurised water reactors, to understand their properties and in-service behaviour under thermo-mechanical and chemical stresses and under strong irradiation.

Who should take this course ?

  • Qualified engineers, scientists and technicians in charge of fabrication, characterization, application and safety evaluations of Zr based components for nuclear purposes.
  • Students carrying out specialised studies on materials science and nuclear engineering.

Learning outcomes

Acquire a general view of Zr alloys from the processing to in service properties including safety concerns:

  1. Highlight the main processing parameters affecting the as-received material properties,
  2. Explain the relationship between the microstructure evolution and the physico-chemical and mechanical properties: under irradiation, during corrosion, oxidation and hydriding in light water reactors environment, under accidental scenarii,
  3.  Give a reactor feedback and next future trends.
  • Lectures are given by academic and professional experts  of recognized standing in Material Sciences.
  • Practical work on actual fuel cladding specimen after thermal treatments and with or without chrome plating (SEM observations)
  • History and overview of Zr alloys for nuclear applications.
  • Processing and forming of industrial components.
  • Phase diagrams (includes Zr-H and Zr-O) and control of microstructures (in Zry and Zr-Nb).
  • Anisotropy, deformation mechanisms, texture development, mechanical properties.
  • Irradiation effects. Effects on microstructure. Creep and growth.
  • Mechanical behaviour after irradiation.
  • Corrosion in water (without and under irradiation).
  • High temperature oxidation and LOCA behaviour.
  • Impact of H Pick-up: embrittlement, RIA, post irradiation creep.
  • Laboratory work on fractography & metallography of Zr claddings.
  • Reactor feedback and future trends in design and requirements.
  • Enhanced-Accident-Tolerant-Fuels » (E-ATF) coated Zr claddings

Basic skills in material science are required

Auto-evaluation and final MCQ

Teaching method and tools

Contact(s)

Program manager :

Bertrand REYNIER
bertrand.reynier@cea.fr
+33 1 69 08 49 75

training sessions

No training session is scheduled for the moment, if this training interests you, please contact us.

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