Mechanistic Understanding of Irradiation-induced Corrosion of Zirconium Alloys in Nuclear Power Plants


Book Description

Failures in the basic materials used in nuclear power plants continue to be costly and insidious, despite increasing industry vigilance to catch failures before they degrade safety. For instance, the overall costs to the US industry from materials problems could amount to as much as $10 billion annually. Moreover, estimates indicate that the cost of a pipe failure in a nuclear plant is one hundred times greater than the cost of a similar failure in a coal-fired plant. There are important practical stimuli and much scope for further understanding of the effects of irradiation on Zr-alloys (and other materials used in nuclear installations) by careful experimentation. Moreover, these studies need to address the effect of irradiation on all components of heterogeneous systems: the metal, the oxide and the environment, and especially those processes recurring at the interphases between these components. The present paper is aimed at providing specialists with some systematic information on the subject and with important considerations on the key items for further experimentation.




Transmission Electron Microscopy Characterization of Zircaloy-4 Subjected to Ion Irradiation


Book Description

In the operation of a nuclear reactor, the performance of the fuel cladding is critical to ensuring safe and reliable operation of the reactor. The current generation of Light Water Reactors utilizes claddings made from zirconium alloys. The material used for nuclear reactors must be able to withstand temperatures above 3000C while also being exposed to water, high pressures, and radiation. During operation, the zirconium cladding corrodes and picks up hydrogen which can adversely affect its performance. The corrosion mechanisms at work have yet to be fully characterized, especially the influence of irradiation. In order to better understand the mechanisms at work and characterize the behavior of zirconium alloys under reactor conditions, the Mechanistic Understanding of Zirconium Alloy Corrosion (MUZIC) consortium focused on the autoclave corrosion (MUZIC-1) and hydrogen pickup (MUZIC-2) outside of irradiation. The MUZIC-3 effort focuses on corrosion under irradiation. While it would be optimal to test reactor-irradiated samples, the difficulties posed by irradiating, corrosion testing, and examining these samples makes ion irradiation a more appealing manner of irradiation. Using doses and temperatures adjusted for substitution of protons for neutron radiation, this experiment seeks to characterize the effects of irradiation on the base metal, oxide layer, and water, both separately and jointly, on the corrosion of zirconium alloys. In this thesis, the beginning stages of this project, part of MUZIC-3, are presented. This involves verification of the effect of proton irradiation (which is used to represent neutron irradiation) on the base metal and characterization of the irradiated samples. The corrosion testing of this irradiated material will provide a reference for the effect of irradiation induced microstructure changes to the base metal on corrosion. In order to characterize the samples, chemical analyses and observations on crystallinity of secondary phase particles are needed. Along with the analysis of second-phase precipitates, assessment of dislocation loops to observe similarities between different radiation types is also required. Accordingly, samples were irradiated with charged particles (protons and zirconium ions) at the Michigan Ion Beam Laboratory and focused ion beam samples were prepared for transmission electron microscopy examination. The microstructure of the base metal is examined for a range of doses and irradiation temperatures and compared to the microstructure created under neutron irradiation as a preliminary to corrosion testing of irradiated samples. The results are discussed in light of existing literature.




Mechanistic Understanding of Zirconium Alloy Fuel Cladding Performance


Book Description

A review is presented of work performed in our group over the years in the areas of radiation damage, corrosion, hydrogen pickup, hydriding, and the mechanical behavior of zirconium alloy nuclear fuel cladding with the goal of developing a greater mechanistic understanding of cladding performance in service.




Zirconium in the Nuclear Industry


Book Description

The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr




Zirconium in the Nuclear Industry


Book Description




Zirconium in the Nuclear Industry


Book Description







Nuclear Corrosion


Book Description

Nuclear Corrosion: Research, Progress and Challenges, part of the "Green Book” series of the EFC, builds upon the foundations of the very first book published in this series in 1989 ("Number 1 - Corrosion in the Nuclear Industry”). This newest volume provides an overview on state-of-the-art research in some of the most important areas of nuclear corrosion. Chapters covered include aging phenomena in light water reactors, reprocessing plants, nuclear waste disposal, and supercritical water and liquid metal systems. This book will be a vital resource for both researchers and engineers working within the nuclear field in both academic and industrial environments. Discusses industry related aspects of materials in nuclear power generation and how these materials react with the environment Provides comprehensive coverage of the topic as written by noted experts in the field Includes coverage of nuclear waste corrosion




Irradiation Induced Redstribution of Alloying Elements in Zr-Nb Alloys and Its Effect on Corrosion Kinetics


Book Description

Zirconium-based alloys have been used in nuclear fission reactors, because of their low thermal neutron cross-section, good mechanical strength, and adequate corrosion resistance. In pressurized water reactors, one of the major reasons that Zr-Nb alloys have widely replaced Zircaloys-4 is the absence of the accelerated oxide growth at high burnup. Although such great advantages have led to the development of many advanced commercial Zr-Nb alloys, the reasons behind the enhanced in-reactor corrosion resistance are still unclear. The distribution and concentration of alloying elements in the substrate have been suspected bo play a major role in controlling in-reactor corrosion kinetics. The major hypothesis being tested in this thesis study is that the enhanced corrosion resistance of irradiated Zr-Nb alloys is a result of irradiation-induced reduction of Nb concentration in solid solution, due to the nucleation and growth of Nb-rich irradiation-induced platelets (IIPs)/nanoclusters. To validate our hypothesis, a systematic study has been performed to understand the irradiation induced microstructure and microchemistry evolution and the subsequent effect on the corrosion kinetics of Zr-xNb (x=0.2, 0,4, 0,5, 1.0) model alloys. The microstructure and microchemistry of as-received materials were characterized under STEM/EDS/APT. Then, 2 MeV proton irradiation has been performed on these model alloys at 350 °C up to 1 dpa. (S)TEM/EDS has been used to study the size and density evolution of the native precipitate and IIPs as a function of irradiation dose. The major use of APT is to quantify the Nb concentration in the solid solution as a function of irradiation dose in order to support our hypothesis. The IIPs crystal structure and growth mechanism have been particularly inverstigated using HRSTEM and 4D-STEM. After the characterization, the irradiated materials were corroded in autoclave to study if the proton irradiation leads to subsequent lower corrosion rate. Lastly, the same characterization techniques and methods have been used to study neutron irradiated commercial alloys, M5®, ZIRLO® and X2®, in an effort to compare the results with proton irradiation. The possible IIPs nucleation and growth mechanism and the effects of irradiation-induced Nb redistribution on the corrosion kinetics are the major focuses of the discussion section.




Zirconium in the Nuclear Industry


Book Description

Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.