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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

Nuclear fuel cycle


Materials for nuclear


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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


Program manager :

Bertrand REYNIER
+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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