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.




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







Damage Stucture in Zirconium Alloys Irradiated at 573 to 923°K. [Neutron Fluence 1 X 1025 N. M−2].


Book Description

The microstructures of annealed zirconium, Zircaloy-2 and Zr-2.5 wt % Nb alloy and of Zr-2.5 Nb containing .cap alpha.' were studied after neutron irradiation to fluences approximately equal to 1 x 1025 n x m−2 in the temperature range 573 to 923°K. The principal form of damage was dislocation loops which increased in size and decreased in density with increasing temperature and which did not exist above 773°K. The Burgers vector of the loops was consistent with a/3 1120. Half or more of the loops were of vacancy type. No dislocation networks or voids were seen. It is argued that the bias of loops for self-interstitial atoms in .cap alpha.-zirconium is very weak, permitting competitive vacancy and interstitial loops, preventing growth of loops into gross dislocation structure, and depressing the vacancy super-saturation so that voids cannot arise.




Comprehensive Nuclear Materials


Book Description

Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field







Effect of Stress on Radiation Damage in Neutron Irradiated Zirconium Alloys


Book Description

Structures developed in zirconium alloys during irradiation creep have been characterized by transmission electron microscopy (TEM). Alloys investigated were annealed Zircaloy-2, cold-worked Zircaloy-2 and cold-worked Zr-2.5Nb pressure tube material. Thin films were taken from material deformed in the NRU, NRX and Pickering-3 reactors at temperatures of 530 to 600 K under stresses of 117 to 552 MPa giving strains in the range 0.14 to 8.8 percent. Stress-induced orientation of dislocation loops makes a negligible contribution to irradiation creep at all stresses. At the lower stresses (and hence strains), the size and distribution of the damage is unaffected by stress, being the same in the head and gage sections of creep specimens. At higher stresses (strains), there is much clearing of the damage by plastic deformation. The deformation however is very uneven, producing structures in different grains of the same specimen that can show no deformation, swaths cleared of irradiation damage, or dislocation tangles or cell formation. The relevance of these TEM observations to irradiation creep mechanisms is discussed.







Physics Briefs


Book Description