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Fuel Reliability Assessment Through Radiochemistry and Poolside Examinations (LCC20) – 2 Reports

BWR Chemistry, Coolant Chemistry & Corrosion, PWR/VVER Chemistry

Our latest two-volume report, Key Emerging Issues and Recent Progress in PWR/VVER, BWR and New Reactors, provides a comprehensive technical review of the most important operational experience, research findings, and emerging trends presented at the NPC-2025 International Conference in Busan, Korea. The reports cover advanced developments in primary and secondary water chemistry, fuel performance, corrosion mitigation, source term reduction, shutdown chemistry, flexible operation, aging management, radwaste treatment, SMRs, advanced reactors, and digital chemistry optimization programs.

Volume I focuses on PWR/VVER primary and secondary system chemistry, including fuel reliability, CRUD behaviour, zinc injection, potassium hydroxide implementation, steam generator degradation, hydrazine alternatives, shutdown chemistry, and source term control strategies. The report also includes detailed operating experience from EPR startups, chemistry optimization for long-term operation, and extensive benchmarking of global reactor fleets.

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Volume II covers BWR chemistry and radiochemistry, auxiliary and cooling water systems, radwaste treatment technologies, flexible operation, chemistry compliance management, advanced reactor and SMR developments, and long-term aging management. It includes practical lessons learned from recent plant restarts, innovative decontamination technologies, AI-driven chemistry monitoring, and new approaches for minimizing radiation exposure and extending plant lifetime.

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Together, the reports provide utilities, regulators, vendors, and technical specialists with a unique consolidated reference of the latest international developments in nuclear plant chemistry and operational performance, supported by detailed technical analysis, industry data, and real plant experience from operating nuclear fleets worldwide.

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Pellet Cladding Interaction, PCI, in PWRs/VVERs/BWRs (ZIRAT30/IZNA25)

Fuel Material, Mechanical Properties

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

Fuel Material, Interim Dry Storage

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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Reduction of Out-of-Core Radiation Fields and the SFAIRP Principle

Coolant Chemistry & Corrosion, PWR/VVER Chemistry

This report explains how the right choice of materials, chemistry control, and design decisions can significantly cut radiation fields caused by isotopes like Co-60, Co-58, and Sb-124. With real-world lessons from Angra 2 and European reactors, it shows how to apply the SFAIRP principle (So Far As Is Reasonably Practicable) to reduce dose rates, improve outage performance, and extend plant lifetimes.

Engineers, operators, and regulators will benefit from clear strategies, benchmark data, and the introduction of a new Ultimate ALARA Index (UAI) that links energy production with radiation safety.

HOW TO GET MORE INFORMATION, E.G: PRICE AND HOW TO ORDER: sales@antinternational.com

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Interim Dry Storage – Creep (ZIRAT29/IZNA24)

Fuel Material, Interim Dry Storage

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)

Fuel Material, LOCA and RIA Fuel Performance

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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Fuel Reliability Assessment Through Radiochemistry and Poolside Examinations (LCC20)

Coolant Chemistry & Corrosion, PWR/VVER Chemistry

Nuclear power plants are capable of operating safely with leaking fuel in their cores and have done so on many occasions. Fuel failures affect three principal areas:  radiological safety, core and plant operation, and costs. With increased fuel reliability, the effects of leaking fuel on each these areas are heightened. The issue of secondary degradation is particularly severe because of the large increases in radiation exposure that such events can cause in the primary coolant and clean-up systems.  

The use of radiochemistry for the detection and tracking of leaking fuel will obviously not prevent failures from occurring, but can provide information for mitigating the radiation exposure, operational and economic consequences. The objective of this report is to identify methods for detecting failed fuel and assessing conditions associated with leaking rod(s) during operation in an NPP.  The report provides background for understanding the capabilities and limitations of the tracking and assessment methods. 

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Zinc Chemistry in BWRs, PWRs and VVERs (LCC20)

Coolant Chemistry & Corrosion, PWR/VVER Chemistry

This report describes the use of zinc injection technology in BWRs, PWRs and VVERs plants worldwide. Zinc addition is a mature technology applied both in PWR and BWR reactors while its implementation in VVER plants is still waiting to the commissioning of new reactor designs. It has been a crucial milestone in coolant chemistry for controlling and reducing radiation build-up, operational radiation as well as IGSCC mitigation in BWRs and mitigation of Primary Water Stress Corrosion Cracking (PWSCC).

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Key Emerging Issues and Recent Progress Related to Plant Chemistry/Corrosion (PWR, CANDU, and BWR Nuclear Power Plants) (LCC19)

Coolant Chemistry & Corrosion, PWR/VVER Chemistry

Safety and reliability of power plants are becoming increasingly important factors since many plants are aging and have obtained license renewal for continued power operation and also for new reactors using different technologies that are or will be in design, construction, commissioning, or start-up stage. Therefore, sharing plant operating experiences, sharing lessons learned, and sharing new industry research are all crucial in order to maintain the nuclear power plant fleet in a healthy condition as well as for new reactors using different technologies that are or will be in design, construction, commissioning or start-up stages.

This report on Key Emerging Issues and Recent Progress, ANT International has collected the most relevant experiences and advanced research exposed at the Nuclear Plant Chemistry Conference NPC-2023 that took place in Antibes Juan-les Pins, France in September.

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PWRs Operation and Maintenance Raw Water Systems (LCC19)

Degradation of Different Types of Structural Materials, Structural Material Degradation

Raw water has a major safety role as acting as cold source for plants. Raw water is used for:

  • Cooling the condenser, either in open or in closed circuits;
  • Providing water for service water systems;
  • Providing water to the Fire Fighting System;
  • Providing water to the Auxiliary Feedwater Tank in case of emergency (earlier units).

The report covers the following topics: design consideration, raw water chemical treatments, operating experience along with the maintenance programmes of raw water systems.

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