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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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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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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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