EPRI LWRs Material Reliability 2016 Conference

(LCC13 AR)

For many years, EPRI has organised conferences on light water reactors materials reliability. Given there was neither Environmental Degradation nor Fontevraud conferences in 2016, EPRI took the opportunity of providing this conference in 2016. During days 2 to 4, 104 slides were presented in 3 parallel sessions, which covered 18 topics on a whole range of concerns of LWRs.

This ANT report is a summary of some of the most important slides, with little text added. The report contains a summary and a chapter dedicated to the relevance of some results presented for the Industry. This conference brings up-to-date information regarding the latest research, plant experiences, analyses, planning and solutions for increased materials reliability in BWR and PWR components. The report is of interest for engineers in charge of expertise, materials, chemistry, non-destructive testing, fracture mechanics, corrosion, irradiation effects, degradation mitigation, and modelling.

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Key Emerging Issues and Recent Progress related to Structural Materials Degradation (PWRs, VVERs, CANDUs and BWRs)

(LCC12)

During operation, the materials used for the construction of components react with light water reactor environment and cause component degradation, including cracking at welds and piping. Such degradation is due to irradiation, corrosion, fatigue, and other damage mechanisms, and has remained a severe opera­tional challenge for utilities. Details on such degradation are reg­ularly reported and published in scientific journals and at utility workshops and conferences. Foremost in the latter category are those organized since 1983 by NACE, TMS, ANS and CNS in the Environmental Degradation of Water Reactor Materials.

This Report contain highlights from the 17th Inter­national Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, which was held in Ottawa, Canada in August 2015. It covers PWRs, VVERs, CANDUs and BWRs. Over 150 papers were published at this conference and this Report cover diverse topics touching on specific degradation modes in various alloys, such as:

• PWSCC of cold worked Alloy 690 and its weld metals
• Corrosion fatigue of stainless steels in PWRs and BWRs
• Fuel cladding materials
• Cracking of Alloy 718 and X750
• Flow assisted corrosion
• Oxide films and characterisation in PWR Secondary systems
• SCC of Alloy 82 and 182 welds
• IASCC of stainless steels
• Irradiation studies involving ion and neutron irradiation
• Synergistic effects of thermal ageing and neutron embrittlement
• Alloy 600 oxidation and mechanisms in PWRs
• PWR Field experience
• Advances in in-situ monitoring and ex-plant and mockup component evaluation
• Generation IV materials research
• Fundamental studies involving new state-of-the-art microscopy and other techniques
• Fukushima accident related research on SS corrosion in sea water
• Dry Cask Storage System and Waste Container Corrosion

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Environmentally-Assisted Degradation of Structural Materials in Water Cooled Nuclear Reactors – An Introduction

(SMDR)
This Report is an updated and expanded version of the 2006 ANT International Report entitled: “Environmentally-assisted degradation of structural materials in water cooled nuclear reactors” authored by Dr. Peter Ford.

The objective of this Report are twofold: first, to provide an updated edition of the 2006 Report cited above and, second, to provide a text-book that complement the 4-day webinar on environmentally-assisted degradation of structural materials.

The Report is intended for people new to the subject, or who need a “refresher” on the essential factors behind component failures and the subsequent mitigation actions. Such a focus is critical at this time, given the ongoing retirement of experienced personnel and the loss of “corporate memory” relating to the management of materials degradation. This loss is being felt in areas of reactor license renewal, power uprates, load following, and the certification and construction of advanced designs of both BWRs and PWRs.

In the same series as this updated and expanded Report, there are four other detailed Reports that analyse the behaviour of structural material degradation of various alloys commonly used in Pressurised Water Reactors (PWR/VVER) and Boiling Water Reactors (BWR).

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