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.
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.
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.
(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.
(ZIRAT23/IZNA18 AR)
The overall objective of the ZIRAT Programme is to enable the nuclear utilities and laboratories to:
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.
(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.
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.
(ZIRAT22/IZNA17 AR)
The overall objective of the ZIRAT Programme is to enable the nuclear utilities and laboratories to:
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.
(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.
(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.
