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View allAccident Tolerant FuelInterim Dry StorageManufacturingAnnual Reports Presenting Key Results Published During 18 MonthsOther IssuesCorrosion and Hydriding of Zr AlloysProperties of Hydrides and Impact on Fuel In-Reactor PerformanceDimensional Changes of Fuel Assemblies/Channels and ComponentsMechanical PropertiesLOCA and RIA Fuel PerformanceFuel (UO2 and MOX) PerformanceFuel ReliabilityMicrostructure of Zr Alloys and Effects on Performance
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Pellet Cladding Interaction, PCI, in PWRs/VVERs/BWRs (ZIRAT30/IZNA25)

The Special Topic Report on PCI-SCC provides a comprehensive state-of-the-art review of Pellet-Cladding Interaction assisted by Stress Corrosion Cracking (PCI-SCC), one of the most important fuel reliability and operational flexibility challenges in modern nuclear power plants. The report consolidates decades of international R&D, operating experience, experimental results, and recent advances in modeling and fuel design related to PCI-SCC.

The report covers the fundamental mechanisms behind PCI-SCC, laboratory testing techniques, in-reactor experience, fuel and cladding design improvements, advanced modeling approaches, and operational methodologies for flexible power operation and load-following conditions. Special attention is given to modern fuel designs, fuel conditioning and deconditioning strategies, Accident Tolerant Fuel (ATF) developments, and the increasing importance of PCI-SCC under evolving grid demands and extended reduced power operation (ERPO).

The report also highlights recent progress in mechanistic understanding, post-irradiation examination techniques, simulation tools, manufacturing quality improvements, and operational guidelines used internationally to manage PCI-SCC risks.
The report provides utilities, fuel vendors, regulators, and technical specialists with an up-to-date technical reference on PCI-SCC, combining operating experience, recent research, advanced modelling developments, and practical operational guidance to support improved fuel reliability, operational flexibility, and informed decision-making for current and future reactor operation.

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Interim Dry Storage – Delayed Hydride Cracking

ZIRAT30 Special Topic Report on Delayed Hydride Cracking (DHC) provides a comprehensive state-of-the-art review of hydrogen-related degradation mechanisms in zirconium alloys, with particular focus on delayed hydride cracking during interim dry storage of spent nuclear fuel. The report consolidates decades of international operating experience, experimental studies, fracture mechanics evaluations, and recent advances in mechanistic understanding related to DHC and hydride behaviour in zirconium alloy fuel claddings and pressure tubes.

The report covers the fundamental behaviour of hydrogen in zirconium alloys, including hydrogen solubility, diffusion, hydride precipitation, hydride fracture properties, and the influence of stress, temperature, irradiation, alloy composition, and texture on DHC susceptibility. Extensive operating experience from BWR, PWR, VVER, CANDU, and RBMK reactors is reviewed together with detailed discussions on crack initiation, crack growth mechanisms, threshold stress intensity factors (KIH), and dry storage performance assessments.

Special attention is given to recent international research on DHC mechanisms, including advanced experimental techniques, neutron radiography, finite-element modelling, hydride phase characterization, and the evolving understanding of diffusion-controlled crack growth models. The report also evaluates the impact of irradiation hardening, liner concepts, hydride reorientation, temperature history, and material microstructure on DHC behaviour under storage-relevant conditions.

The report provides utilities, regulators, fuel vendors, and technical specialists with an up-to-date technical reference on delayed hydride cracking, supporting informed decision-making related to spent fuel dry storage, fuel integrity assessments, long-term storage safety, and zirconium alloy performance under extended storage conditions.

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Interim Dry Storage – Creep (ZIRAT29/IZNA24)

Dry storage of commercial spent nuclear fuel (CSNF) is a well-established technology. As of 2024, spent nuclear fuel elements from commercial power plants and from research reactors have been stored in a dry state for nearly 40 years and 50 years, respectively. The overarching goal is preventing CSNF degradation that would result in multiple fuel rod failures during dry storage, at least through post-storage retrieval for reprocessing or placement and sealing in a container at a final disposal repository. Since performance with dry storage has already been shown acceptable for tens of years, the emphasis is presently on any changes that may occur during extensions of the dry-storage time periods to a hundred years. This Special Topic Report addresses Thermal Creep. It is shown that thermal creep under normal conditions of storage is unlikely to result in cladding rupture for present, as well as later.

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Loss of Coolant Accident (ZIRAT29/IZNA24)

During and after a Loss Of Coolant Accident, the core must be coolable, and structural integrity must be maintained. The fuel cladding temperatures during a LOCA may increase over 1 000°C, resulting in:

• Steam oxidation of the cladding
• Cladding embrittlement,
• Ballooning and burst if the rod internal pressure is large enough and,
• Finally, if the high temperatures are maintained long enough, the Zr oxidation reaction may become so exothermic that fuel rod cannot be cooled.

Fuel safety criteria in most countries are based on the United States Nuclear Regulatory Commission (USNRC) criteria. The Japanese and the French criterion are not based on zero ductility of cladding, but on the failure threshold value determined in the integral thermal shock tests under restrained conditions. The idea of this report is to tell the “LOCA story”, from physics to acceptance criteria including a short background to the LOCA phenomenon focusing on the data and information which are used in the current design approaches focusing on BWRs, PWRs, VVERs.

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STR on Interim Dry Storage of Commercial Spent Nuclear Fuel – An Update (ZIRAT28/IZNA 23)

Except for a few countries (Finland, Sweden, France, and possibly Canada), the timing for establishing a geologic repository has been shown to be unpredictable. Therefore, spent fuel storage will remain the last backend operation for the foreseeable future in many countries. With proper attention, the radiological impact of storage is very low, but regulatory agencies have placed a heavy burden on licensees because of concerns related to the highly negative public perception related to the presence of spent fuel storage facilities in our biological environment. Therefore, locations where spent nuclear fuel (SNF) is or will be stored and their chosen storage technologies are the subjects of much scrutiny.

The focus of this 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. More specifically, the following topics are treated in the report: 

  • Update of “Back-end” issues
  • Thermal creep behaviour in relation to hydride reorientation
  • PWR fuel rod cladding failure due to the hydrogen migration in spent fuel
  • Update on any work on storage, transportation, long term issues
  • Correlation between cooling rate and hydride reorientation. In particular, the case of fast cooling when the cask containing SNF is flooded with water, from a cladding temperature of ~350°C to ~30°C, is examined.

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Accident Tolerant Fuel: an update report 2020-2023 (ZIRAT28/IZNA 23)

During the last decade the development of various Accident Tolerant Fuel (ATF) concepts has come into focus of both research and industry communities in the USA, Europe and Asia. The accident tolerant fuel program is aiming towards improving the safety of nuclear energy by investigating materials that can replace or modify the current uranium-dioxide nuclear fuel and zirconium-based cladding. This research programs are being supported by all major nuclear countries since 2011.

This updated version of the previously issued ANT International 2021 ATF report will provide the reader with a useful, quick but comprehensive overview of the latest ATF technical developments. It will give relevant information about ATF cladding and fuel, appropriate warnings and useful insights to nuclear fuel engineers and designers as well as to the fuel buyers.

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Nuclear Fuel Behaviour Under RIA Conditions

(ZIRAT27/IZNA22)

The content of the Updated RIA is basically the same as the original report (Nuclear Fuel Behaviour under RIA conditions published within the ZIRAT21/IZNA Programmes), see information below. 

The main focus of this report is to give an update on two major subjects: (1) new RIA tests and the interpretation of the results and (2) new RIA related regulations. New RIA tests will complete the existing data base and their potential significance for RIA modelling or for RIA ruling. New acceptance criteria for RIA issued by US NRC are briefly described as well as some national approaches, different from the US NRC

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Pellet-Cladding Interaction (PCI and PCMI): progress in nuclear industry

(ZIRAT27/IZNA22)

Pellet-Cladding Interaction (PCI) and Pellet-Cladding Mechanical Interaction (PCMI) remains an important phenomenon in the modern nuclear reactor fuel engineering. This topic was addressed in detail in ZIRAT-11 Special topical report on PCI issued in 2006. Since then, a substantial development has taken place, such as introduction of new cladding materials and additive fuel pellets. In addition to this, flexible power operation is being considered by many utilities worldwide. This introduces new challenges to the safe and reliable operation of the nuclear fuel. The report discusses up-to-date developments in the above mentioned areas and summarizes the main outcomes of the R&D work performed since 2006.

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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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