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ANTIA Seminars

ANT International Academy provides a variety of Seminars in the areas of Fuel Material, Structural Material Degradation, and Coolant Chemistry and Corrosion. There are three levels of depth:

  • Basic
  • Intermediate ▪▪
  • In-depth ▪▪▪
  • Expert ▪▪▪▪

The Seminars are held annually in Spain during the month of March. The length is one to four days. Before the Seminar, the presentation material will be made available to the participants. After the Seminar, a certificate of seminar attendance will be issued by ANT International to the participants.





Scientific backgrounds for the behaviour of Zr alloys in nuclear reactor ▪▪▪▪

This seminar is specially designed for R&D scientists concerned by the scientific aspects of Zr alloys development or behaviour, interested in the rising techniques or the advanced approaches currently used for other alloys or just under the early stage of consideration for Zr alloys.

Considering the specific properties of Zr alloys, we will review major advanced basic material science approaches and techniques, source of recent progresses in other alloys, with the aim of rising areas or techniques of potential fruitful research for Zr alloys.

In this seminar, we will cover the fields of microstructure from a fundamentals point of view, with special emphasis on phase equilibria, free energy (at phase or atomic scale), kinetics and phase transformations either by precipitation or bainitic (β => α). A discussion will be given on the modifications induced by irradiation and the new ways to handle it. This seminar is aimed at being the first step of a series to be continued the next two years with the additional topics, with the same scientific approach : Mechanical properties and irradiation induced changes, followed by Interaction between water and Zr alloys in reactors – corrosion, H pick-up.

THE FOLLOWING TOPICS ARE COVERED:

  • Need to control microstructure of Zr alloys
  • Classical approach : Thermodynamics and kinetics
  • Computational Material Science
  • Transfer of scientific techniques or concepts
  • Prospects

DETAILED INFORMATION:

LECTURERS:

Dr. Clément Lemaignan

DATE AND LOCATION:

March 2017, Mallorca

DURATION:

4 hours

ASSOCIATED LITERATURE:

Not available


Delayed Hydride Cracking in Zirconium Alloys ▪▪▪

Delayed Hydride Cracking (DHC) has been responsible for failures of components in nuclear reactors and chemical plants made from zirconium alloys. This type of cracking is caused by fracture of hydrides accumulated by the movement and concentration of hydrogen to the ends of gradients in stress, temperature and alloy composition. To guard against this mechanism of cracking during the lifetime of components, one needs knowledge of hydrogen concentration, stress, and distribution of flaws and temperatures. In this seminar the features of the various failures by this mechanism will be described; the relative importance of variables such as hydrogen concentration, stress intensity factor, temperature history and microstructure will be outlined; models for DHC will be described and methods for avoiding DHC during normal reactor operation and during storage of spent nuclear fuel will be described so the lessons learnt can be applied. This topic is important because zirconium components contain hydrogen and tensile stresses and operators need to evaluate the proximity of their components to the limits of DHC.

THE FOLLOWING TOPICS ARE COVERED:

  • Introduction to Delayed Hydride Cracking (DHC)
  • Component failure by DHC
  • Hydrogen in zirconium alloys
  • Hydride properties
  • Basic mechanism of DHC
  • Implications of mechanism for behaviour of a crack
  • Experimental methods
  • Phenomenology and dependencies on:
    • Models of crack growth by DHC
    • Implications for structural integrity

DETAILED INFORMATION:

LECTURERS:

Dr. Kit Coleman

DATE AND LOCATION:

February 2017, Clearwater Beach, FL., USA

DURATION:

1 day

ASSOCIATED LITERATURE:

Not available


Thermal/Hydraulic in PWR Fuel

The design of this ½ day seminar is to cover the basic physical laws related to nuclear fuel and show how these directly affect the design, analysis, and operation of both the fuel and related reactor systems. The discussion will show the links from the fundamental equations to the fuel design, through the cycle analyses, and culminating with the operating limits in PWRs. The training will focus mainly on the thermal-hydraulic and neutronics of the fuel but will also include key fuel rod mechanical items. In addition, a short discussion will show real life examples of how application of basic fundamentals can explain observed operating anomalies.

THE FOLLOWING TOPICS ARE COVERED:

  • Neutronics
    • Basic Equations, including key assumptions for PWR application
    • Relationship of key reactivity components to fuel design
    • Safety Criteria for Cycle Operation
    • Key parameters for Fuel Design, Cycle Design, and Required Cycle Analyses
  • Thermal Hydraulics
    • Basic Equations, including key assumptions for PWR application
    • Forms of Heat Transfer in PWRs
    • Safety Criteria for Cycle Operation
    • Key parameters for Fuel Design, Cycle Design and Required Cycle Analyses
  • Fuel Rod Mechanical
    • Basic Equations, including key assumptions for PWR application
    • Safety Criteria for Cycle Operatio
    • Application of Fundamentals to Operating Anomalies
    • Two examples of how understanding the basic physics of the system can help in identifying and explaining unexpected operating anomalies

DETAILED INFORMATION:

LECTURERS:

Mr. Kenneth Epperson

DATE AND LOCATION:

March 2017, Mallorca

DURATION:

4 hours

ASSOCIATED LITERATURE:

Not available


Dry Storage of Commercial Spent Nuclear Fuel ▪▪▪

Interim storage of spent fuel will be required until sufficient capacity in permanent geologic repositories or until more advanced technology options become available. Spent fuel is first stored in pools (ponds) located within the nuclear power plant facilities. Given the limited capacity of these installations, additional storage capacity located at either centralized or reactor site facilities are required. In this seminar, the focus will be on the performance of commercial LWR fuel assembly components, with emphasis on Zircaloy-based alloy cladding, during long-term storage of the spent fuel in a dry, inert environment such as helium. Potential degradation mechanisms of cladding alloys will be examined under normal and offset conditions of storage. Changes in cladding mechanical properties will be reviewed in order to properly assess the impact of interim storage upon subsequent spent-fuel management activities, such as transportation. The international experience with dry storage technology implementation and regulatory positions will frame the discussions. Participants will be invited to contribute their own experience and share issues that they have dealt, or are dealing with in satisfying their country-specific regulatory requirements.

THE FOLLOWING TOPICS ARE COVERED:

  1. Interim Storage of Spent Fuel and HLW: A Global Activity
  2. Commercial Spent LWR Fuel: Brief technology and nomenclature review
  3. Dry Storage Technology
  4. Dry Storage’s Defining Characteristic: Thermal history during fuel transfer and dry storage operations
  5. Potential Fuel Degradation Mechanisms During Dry Storage
  6. Dry Storage Wrap-up
  7. Transportation
  8. Seminar Evaluation

DETAILED INFORMATION:

LECTURERS:

Dr. Albert Machiels and Mr. Friedrich Garzarolli

DATE AND LOCATION:

March 2018, Mallorca

DURATION:

2 days

ASSOCIATED LITERATURE:

Dry Storage Handbook




Maintaining Nuclear Reactor Components Performance and Integrity for Plant Safety

The 3-day seminar follows previous seminars offered by ANTI International on aspects related to degradation experienced by nuclear reactor components. The objective of the 3-day Seminar is to give an overview of the world wide operational experience with major SCCs in nuclear power plants (PWR, BWR). Herby various examples of safety-related reactor components are presented suffering degradation due to specific ageing mechanisms. Measures to prevent failure and precautions to maintain component performance and integrity to ensure long term operation are presented.

The purpose of the seminar is to:

  1. Enhance the awareness and provide practical information relevant to implement component monitoring and addressing ageing in nuclear plant operation, including the current status of non-destructive examination methods used;
  2. Provide practical information on previously experienced ageing of components including, but not limited to, reactor pressure vessel (RPV), piping, nozzles, steam generator tubes, and dissimilar metal welds that accompany such components and others; and,
  3. Present mitigating concepts in case of degradation incidents.

THE FOLLOWING TOPICS ARE COVERED:

  • Brief description of the various types of LWRs and of their main components relevant for safety
  • Materials and materials properties
  • Review of Non Destructive Testing technologies used in LWRs
  • WENRA Reactor Safety Reference Level I “Ageing management”
  • Field experience with PWR components
  • Field experience with BWR components
  • Mitigating concepts in case of degradation incidents (BWR)
  • Degradation in LWR components and Mitigation Techniques
  • Operational surveillance and measures to maintain component performance or reduction of dose rate
  • Brittle fracture analysis of RPV in case of PTS (pressurized thermal shock) in PWR
  • Integrity concept for piping systems with corresponding leak and break postulates in LWR to ensure long term operation

DETAILED INFORMATION:

LECTURERS:

Mr. François Cattant and Dr. Ulf Ilg

DATE AND LOCATION:

March 1-3, 2017, Mallorca

DURATION:

3 days

ASSOCIATED LITERATURE:

Not available