哪位有ASTM STP 1354，共享一下，兄弟急用
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哪位有ASTM STP 1354，共享一下，兄弟急用
本帖被 公民 从 标准资料求助板块 移动到本区(2008-01-28)
因做毕业设计，急需用到该会议论文集，如果哪位兄弟有，麻烦共享一下，兄弟真不胜感激。
[ 此贴被公民在2008-01-28 10:36重新编辑 ]
很大呢
900多頁
Zirconium in the Nuclear Industry: Twelfth International Symposium
Sabol GP
Pages: 943 Published: 2000
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Overview
Evolution of Dislocation and Precipitate Structure in Zr Alloys Under Long-Term Irradiation
Averin SA, Panchenko VL, Kozlov AV, Sinelnikov LP, Shishov VN, Nikulina AV
Effect of Long-Term Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes
Hosbons RR, Davies PH, Griffiths M, Sagat S, Coleman CE
Post-Irradiation Characterization of Ultra-High-Fluence Zircaloy-2 Plate
Mahmood ST, Farkas DM, Adamson RB, Etoh Y
Effects of Neutron Irradiation on Microstructure and Properties of Zircaloy
Adamson RB
The Terminal Solid Solubility of Hydrogen in Zirconium Alloys
McMinn A, Darby EC, Schofield JS
The Long-Range Migration of Hydrogen Through Zircaloy in Response to Tensile and Compressive Stress Gradients
Kammenzind BF, Berquist BM, Bajaj R, Kreyns PH, Franklin DG
Influence of Cladding Microstructure on the Low Enthalpy Failures in RIA Simulation Tests
Garde AM
Experiment and Modeling of Advanced Fuel Rod Cladding Behavior Under LOCA Conditions: Alpha-Beta Phase Transformation Kinetics and EDGAR Methodology
Forgeron T, Brachet JC, Barcelo F, Castaing A, Hivroz J, Mardon JP, Bernaudat C
Behavior of Irradiated Zircaloy-4 Fuel Cladding Under Simulated LOCA Conditions
Ozawa M, Takahashi T, Homma T, Goto K
Microstructure and Properties of Zirconium Alloys in the Absence of Irradiation
Charquet D
Influence of Texture on the Fracture Toughness of Unirradiated Zircaloy Cladding
Grigoriev V, Pettersson K, Andersson S
Degraded Fuel Cladding Fractography and Fracture Behavior
Edsinger K, Davies JH, Adamson RB
Understanding Hydrogen in Zirconium
Sawatzky A, Ells CE
Delayed Hydride Cracking in Irradiated Zircaloy Cladding
Efsing P, Pettersson K
Size, Geometry, and Material Effects in Fracture Toughness Testing of Irradiated Zr-2.5Nb Pressure Tube Material
Davies PH, Shewfelt RSW
Failure Mechanisms of Irradiated Zr Alloys Related to PCI: Activated Slip Systems, Localized Strains, and Iodine-Induced Stress Corrosion Cracking
Fregonese M, Régnard C, Rouillon L, Magnin T, Lefebvre F, Lemaignan C
Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes
Bouffioux P, Rupa N
A New Fabrication Process for Zr-Lined Zircaloy-2 Tubing
Abe H, Takeda K, Uehira A, Anada H, Furugen M
Zirconium Alloy Cold Pilgering Process Control by Modeling
Aubin JL, Montmitonnet P, Mulot S
In-Process Investigation of Precipitate Growth in Zirconium Alloys
Herb B, Ruhmann H, König A
Influence of Composition and Fabrication Process on Out-of-Pile and In-Pile Properties of M5 Alloy
Mardon J-P, Charquet D, Senevat J
Irradiation Creep and Growth Behavior, and Microstructural Evolution of Advanced Zr-Base Alloys
Gilbon D, Soniak A, Doriot S, Mardon J-P
Effect of Dilute Alloy Additions of Molybdenum, Niobium, and Vanadium on Zirconium Corrosion
Sabol GP, Comstock RJ, Nayak UP
E110 Alloy Cladding Tube Properties and Their Interrelation with Alloy Structure-Phase Condition and Impurity Content
Shebaldov PV, Peregud MM, Nikulina AV, Bibilashvili YK, Lositski AF, Kuz'menko NV, Belov VI, Novoselov AE
Contribution to a Better Understanding of the Detrimental Role of Hydrogen on the Corrosion Rate of Zircaloy-4 Cladding Materials
Blat M, Legras L, Noel D, Amanrich H
Mechanism of Corrosion Rate Degradation Due to Tin
Takeda K, Anada H
Influence of Transition Elements Fe, Cr, and V on Long-Time Corrosion in PWRs
Broy Y, Garzarolli F, Seibold A, Van Swam LF
Fundamental Study on the Corrosion Mechanism of Zr-0.2Fe, Zr-0.2Cr, and Zr-0.1Fe-0.1Cr Alloys
Murai T, Isobe T, Takizawa Y, Mae Y
Microstructural Aspects of Corrosion and Hydrogen Ingress in Zr-2.5Nb
Urbanic VF, Griffiths M
The Effect of Microstructure on the Corrosion Behavior of Zircaloy-2 in BWRs
Etoh Y, Nonaka Y, Kubo T, Nomata T, Hayashi H, Kitamura M
Impact of Second Phase Particles on BWR Zr-2 Corrosion and Hydriding Performance
Rudling P, Wikmark G, Lehtinen B, Pettersson H
Corrosion of Electron-Irradiated Zr-2.5Nb and Zircaloy-2
Woo OT, McDougall GM, Hutcheon RM, Urbanic VF, Griffiths M, Coleman CE
Evaluation of Zircaloy-2 Cladding Corrosion Characteristics by Simulated BWR Corrosion Loop Test
Shimada S, Asahi K, Nonaka Y, Sasaki M, Kogai T, Hayashi H, Kitamura M, Sakamoto M, Yamawaki M
Irradiation-Enhanced Deformation of Zr-2.5Nb Tubes at High Neutron Fluences
Causey AR, Holt RA, Christodoulou N, Ho ETC
Studies of Zirconium Alloy Corrosion and Hydrogen Uptake During Irradiation
McDougall GM, Urbanic VF, Aarrestad O
Behavior of Lithium and Boron in Irradiated and Unirradiated Oxides Formed on Zircaloy-4 Claddings
Kido T, Wada S, Takahashi T, Uchida H, Komine I, Inoue Y
Contribution to the Understanding of the Effect of the Water Chemistry on the Oxidation Kinetics of Zircaloy-4 Cladding
Pêcheur D, Godlewski J, Peybernès J, Fayette L, Noé M, Frichet A, Kerrec O
Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance
Nanikawa S, Etoh Y, Shimada S, Kubo T, Ito K, Harada H
Electrochemical Examinations in 350°C Water with Respect to the Mechanism of Corrosion-Hydrogen Pickup
Baur K, Garzarolli F, Ruhmann H, Sell H-J
Hydrogen Transport in the Oxide and Hydrogen Pickup by the Metal During Out- and In-Reactor Corrosion of Zr-2.5Nb Pressure Tube Material
Ramasubramanian N, Perovic V, Leger M
High-Fluence Irradiation Growth of Cold-Worked Zr-2.5Nb
Holt RA, Causey AR, Griffiths M, Ho ETC
Stress Distribution Measured by Raman Spectroscopy in Zirconia Films Formed by Oxidation of Zr-Based Alloys
Godlewski J, Bouvier P, Lucazeau G, Fayette L
Role of Intermetallic Precipitates in Hydrogen Transport through Oxide Films on Zircaloy
Hatano Y, Sugisaki M, Kitano K, Hayashi M
Multi-Scale Characterization of the Metal-Oxide Interface of Zirconium Alloys
Bossis P, Lelièvre G, Barberis P, Iltis X, Lefebvre F
Committee: B10
Paper ID: STP1354-EB
DOI: 10.1520/STP1354-EB
可以多次上传的。
STP 1354
Zirconium in the Nuclear
Industry: Twelfth International
Symposium
Pages: 953 Published: 2000
[ 此贴被eden在2008-01-28 07:47重新编辑 ]
附件： ASTM STP 1354-2000.part01.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part02.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part03.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part04.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part05.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part06.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part07.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part08.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part09.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part10.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part11.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part12.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part13.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part14.rar (3907 K) 下载次数:2
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附件： ASTM STP 1354-2000.part15.rar (1268 K) 下载次数:2
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