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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Corrosion and Hydrogen Pickup – 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.

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The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Hydride Reorientation

Hydride orientation has an important effect on fracture toughness of hydride-containing zirconium alloys because hydrides form as approximately linear arrays of platelet-shaped microscopic precipitates with habits on or near the basal planes of the α–Zr matrix in which they form.

This Stand Alone Report (SAR) addresses a key aspect of the issues raised in the foregoing by providing a comprehensive, self-contained and up-to-date review and analyses of the results of studies carried out on the conditions governing hydride orientation in zirconium alloy pressure and fuel cladding tubes used in nuclear reactors. The report combines a detailed theoretical and experimental overview of this subject with the author’s own analyses of these results. These analyses make use of theoretical advances documented in the author’s 2012 book dealing with the effects of hydrogen and hydrides on the integrity of zirconium alloy components. In the author’s 2012 book, emphasis is placed on delayed hydride cracking, which is a localised failure mechanism.

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Covering the Zr Related Results Published During 2016-2017

(ZIRAT22/IZNA17 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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Dimensional Instability and Irradiation Growth of Zirconium Alloys

(ZIRAT22/IZNA17 STR)

In this STR review, aimed specifically at irradiation growth, addressed are conditions of direct interest to LWRs and CANDUs, including information that has mechanistic implications. Irradiation creep was covered earlier by ZIRAT14 Special Topic Report: In-reactor Creep of Zirconium Alloys, authored by Ron Adamson, Friedrich Garzarolli and Charles Patterson, 2009.

The STR addresses all data deemed relevant to understanding irradiation growth, including a broad review of growth mechanisms, and a summary of practical effects of growth on component performance.

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Performance Evaluation of New Advanced Zr Alloys for BWRs and PWRs/VVERs Vol. I

(ZIRAT22/IZNA17)

To meet the current situation with more aggressive reactor environments (higher burnups, changing water chemistries and loading patterns), and resolving fuel performance issues such as BWR channel bowing and PWR assembly bowing, a large number of zirconium alloys have been and are being developed. The main driver for the initial material development in Pressurised Water Reactors (PWRs) has been to reduce corrosion rates and Hydrogen Pick-Up Fractions (HPUFs), which have occasionally limited the maximum discharged burnup.

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