Effects of Recrystallization and Neutron Irradiation on Creep Anisotropy of Zircaloy Cladding


Book Description

Zirconium alloys have hcp crystal structure with low c/a ratio at and below the reactor operating temperatures and exhibit preferred orientations or textures. These textures result in anisotropic mechanical properties, which in-turn affect their in-service behavior such as in-pile creep down of the cladding tubes. The purpose of the present study has been to investigate the effects of recrystallization and neutron irradiation on the anisotropic biaxial creep behavior of Zircaloy cladding tubes. The creep anisotropies of the cold-worked and recrystallized tubes were considered using results from the closed-end internal pressurization tests superimposed with axial loading. The in situ biaxial strain measurements were made using laser and linear variable differential transformer (LVDT) extensometers. Creep data were obtained at various stress ratios, and creep loci were constructed at constant energy dissipation for both cold-worked and recrystallized tubes. X-ray diffraction techniques were used to measure the textures which were then described quantitatively in terms of crystallite orientation distribution functions (CODFs). These CODFs were employed to predict the anisotropy parameters R and P, and the anisotropic creep behavior. The creep behavior of Zircaloy tubes changed with recrystallization. The effect of neutron irradiation on the recrystallized material is modeled by invoking secondary slip systems. The considerable amount of plastic anisotropy observed in the unirradiated recrystallized tubes shows a tendency to decrease and to become isotropic at high fluences. On the other hand, neutron irradiation does not produce any significant changes in the anisotropy of the cold-worked material when radiation growth is taken into consideration.




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




Effects of Radiation on Materials


Book Description

Symposium held in Nashville, Tennessee, June 1990. Almost two-thirds of these 91 papers are authored by researchers outside of the US (including information on research in the former USSR, Japan, and Europe). Topics include: current commercial power reactor systems; microstructural characterization










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







Prediction of Creep Anisotropy in Zircaloy Cladding


Book Description

Due to the hexagonal crystal structure of zirconium and the radial orientation of the basal poles in Zircaloy cladding, the deformation of light water reactor Zircaloy fuel cladding is anisotropic. Plastic deformation of this cladding can be defined by the R and P factors that are the circumferential/radial and axial/radial contractile strain ratios under uniaxial deformation along the axial and circumferential direction, respectively. The in-reactor deformation performance of the cladding can be modeled with good accuracy if the R and P values of the irradiation-induced creep are known.




Journal of Nuclear Science and Technology


Book Description

Includes English language abstracts from Japanese articles in Nihon Genshiryoku Gakkai Shi (Journal of the Atomic Energy Society of Japan)




A Model for Analysis of the Effect of Final Annealing on the In- and Out-of-Reactor Creep Behavior of Zircaloy Cladding


Book Description

The creep behavior of Zircaloy cladding materials depends on materials texture, degree of recrystallization, and chemical composition. This study is devoted mainly to the analysis of the effect of the final annealing (i.e., the degree of recrystallization) on the creep characteristics. For this purpose, data from a series of thermal creep tests are presented and evaluated. In addition, the in-reactor creep data presented by Franklin et al. are used to evaluate the effect of irradiation on cladding creep performance. The out-of-reactor tests are performed under internal pressurization, and the test matrix covers seven conditions with temperatures from 330 to 400°C and hoop stresses between 80 and 160 MPa. Three lots of Zircaloy-2 claddings and one lot of Zircaloy-4 are considered. The difference between the three Zircaloy-2 lots is in their final annealing conditions. The claddings are either stress relief annealed (SRA), recrystallization annealed (RXA), or partially recrystallization annealed (PRXA). The materials used when fabricating the Zircaloy-2 claddings are from the same ingot, and the chemical compositions of the three types of claddings are almost identical. The Zircaloy-4 cladding included in the test is SRA, and the tin content in this material is similar to that in the Zircaloy-2 materials. The creep data are analyzed by separating the primary (transient) and the secondary (steady-state) creep. In this analysis, the Matsuo creep model, which accounts for both primary and secondary creep, is modified, calibrated, and verified using the new thermal creep data. Based on in-reactor data, the thermal creep model is extended to cover also the creep behavior under irradiation. The claddings considered in the in-reactor test were of both SRA and RXA types, and the experiments were made under external pressure. It is observed that for moderate hoop stresses (