Handbooks & Reports

Corrosion and Hydrogen pickup (HPU) – Vol. III

Corrosion and Hydrogen pickup (HPU) mechanisms of Zr alloys remain a top priority of the nuclear industry. Commercial Zr alloys have today adequate in-reactor corrosion properties. However, hydrogen in fuel components limits the fuel performance today during normal operation and accident conditions as well as during transport of spent fuel. Despite more than 50 years research, the corrosion and HPU mechanisms are still not clear. Improved understanding of the in -reactor oxidation and hydrogen pickup mechanisms are thus required. To shed light on theses complicated mechanisms A.N.T. International has published a set of three reports (Vol. I, II and III) with the focus on explaining the very complicated corrosion and hydrogen pickup mechanisms in an understandable manner.

Volume III gives an introduction to the best understood mechanisms of Zr alloys corrosion and HPU mechanisms, with the aim of giving a ”mental image of the phenomena”, more than discussing in detail all the controversial aspects of the current scientific debates.

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Corrosion and Hydrogen pickup (HPU) – Vol. II

Corrosion and Hydrogen pickup (HPU) mechanisms of Zr alloys remain a top priority of the nuclear industry. Commercial Zr alloys have today adequate in-reactor corrosion properties. However, hydrogen in fuel components limits the fuel performance today during normal operation and accident conditions as well as during transport of spent fuel. Despite more than 50 years research, the corrosion and HPU mechanisms are still not clear. Improved understanding of the in -reactor oxidation and hydrogen pickup mechanisms are thus required. To shed light on theses complicated mechanisms A.N.T. International has published a set of three reports (Vol. I, II and III) with the focus on explaining the very complicated corrosion and hydrogen pickup mechanisms in an understandable manner.

Volume II gives more detailed information on:

• Early Zr alloy development
• Out-of-reactor corrosion
• General in Reactor Corrosion and Hydrogen Pickup behaviour
• Effects of alloying impurities on corrosion and HPU
• Corrosion modelling and prediction

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Corrosion and Hydrogen pickup (HPU) – Vol. I

Corrosion and Hydrogen pickup (HPU) mechanisms of Zr alloys remain a top priority of the nuclear industry. Commercial Zr alloys have today adequate in-reactor corrosion properties. However, hydrogen in fuel components limits the fuel performance today during normal operation and accident conditions as well as during transport of spent fuel. Despite more than 50 years research, the corrosion and HPU mechanisms are still not clear. Improved understanding of the in -reactor oxidation and hydrogen pickup mechanisms are thus required. To shed light on theses complicated mechanisms A.N.T. International has published a set of three reports (Vol. I, II and III) with the focus on explaining the very complicated corrosion and hydrogen pickup mechanisms in an understandable manner.

Volume I gives an introduction on the corrosion and hydrogen pickup (HPU) processes in zirconium alloys. The following topics are treated in details:

• The effects of in-reactor irradiation on both zirconium alloys and the coolant (radiolysis)
• Crud sources, transport mechanisms and deposition mechanisms
• Axial offset anomalies
• Reactor cases of severe crud impact on fuel performance

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ACCIDENT TOLERANT FUEL – A REVIEW

The purpose of this report is to present the most promising ATF concepts currently under development at various fuel vendors, research institutions and nuclear laboratories around the world and to provide the reader an independent assessment of the state of the art and potential for development and implementation related with each type of the ATF.

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INTERIM DRY STORAGE OF COMMERCIAL SPENT NUCLEAR FUEL

This Special Topic Report addresses the degradation mechanisms that could potentially affect the performance of spent fuel stored in a dry, inert environment for periods up to ~100 years. The focus of the review is on the spent nuclear fuel rods, and not on the storage system components such as the casks or the canisters and their internal hardware elements.

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FUEL RELIABILITY ASSESSMENT THROUGH TADIOCHEMISTRY AND POOLSIDE EXAMINATIONS

The overall objective of this Special Topic Report (STR) is to provide the knowledge of how the reactor environment (fast neutron flux, temperature, water chemistry, etc.) and the Zr-alloy microstructure, which is a function of material chemistry and manufacturing process, impacts fuel performance during normal operations, transients, design basis accidents and interim dry storage.

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MECHANICAL PROPERTIES

This report is intended to provide basic understanding on the mechanical properties of zirconium alloys, stainless and ferritic steels and nickel alloys. The information can then be used by customers to evaluate their components.

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Zr Alloy Manufacturing and Effects on In-reactor, DBA and Interim Dry Storage Performance

(ZIRAT24/IZNA19 STR)

The overall objective of this Special Topic Report (STR) is to provide the knowledge of how the reactor environment (fast neutron flux, temperature, water chemistry, etc.) and the Zr-alloy microstructure, which is a function of material chemistry and manufacturing process, impacts fuel performance during normal operations, transients, design basis accidents and interim dry storage.

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Covering the Zr Related Results Published During 2017-2018

(ZIRAT23/IZNA18 AR)

The overall objective of the ZIRAT Programme is to enable the nuclear utilities and laboratories to:

• Gain increased understanding of material behaviour related to successful core options for the back end of the fuel cycle.

The objective is met through review and evaluation of the most recent data on zirconium alloys, identification of the most important new information, and discussion of its significance in relation to fuel performance now and in the future. Included in the review are topics on materials research and development, fabrication, component design and in-reactor performance.

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Material Test Reactors and other Irradiation Facilities

(ZIRAT23/IZNA18)

In materials test reactors (MTRs), materials are subject to intense neutron irradiation to study the induced changes. As MTRs are able to reproduce material degradation undergone by materials in power reactors, they provide essential support to the study of ageing of materials in power reactors. MTRs are also being used to irradiate new cladding materials and fuels that are being developed as Accident Tolerant Fuel (ATF) systems.

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