Journal of Nuclear Materials 1996: Vol 229 Table of Contents

2 Pages·1996·0.3 MB·English

Contents
EE A eee ee ae eer are eee er ii Section 2. Steels
Section 1. Zirconium alloys 2.1. Corrosion
Corrosion behaviour of Alloy 800 in high temperature
1.1. Basic properties aqueous solutions: long term autoclave studies,
Point defects diffusion in a-Ti, J.R. Ferndndez, A.M. PE A, AE CEO GEO E PEs « 5 » -
St ite 1 Corrosion behaviour of Alloy 800 in high temperature
Grain boundary diffusion of Onn: in Ze. JL. Azar aqueous solutions: Electrochemical studies, A.M.
M.J. Iribarren and F. Dyment ...... cecil : 10 Olmedo, M. Villegas and M.G. Alvarez ........ 102
Zr diffusion in a-Ti matrices with different Fe content.
Application of models developed to a-Zr self-diffu- 2.2. Surtace activity end treatment
sion, R.A. Pérez, M.L. Aguirre and F. Dyment ... . 15 Cleaning of stainless steel surfaces and oxide dissolu-
tion by malonic and oxalic acids, E.B. Borghi, S.P.
1.2. Phase diagrams Ali, P.J. Morando and M.A. Blesa .........454. 115
ae ; : Activity transport in vessel-type PHWRs, G.A. Urrutia,
Critical evaluation and thermodynamic assessment of A.J.G. Maroto, M. Chocrén and M.A. Blesa ..... 124
the Zr—Pb system, D. Arias, J. Abriata and L.
; Retention of iron and chromium by an anion-exchange
EE bo 6 kee. eke BOs Ob ee ee 24 . cai as
resin under decontamination conditions, H.R. Corti,
Intermetallic phases in the iron-rich region of the D.G. Gémez, E.K. de Blanco, R.A. Jiménez Rebag-
Zr—Fe phase diagram, M.S. Granousky and D. Arias 29
ee es ao a oo beO RS Oe Sw 132
1.3. Hydrogen in Zr Section 3. Corrosion of nuclear materials
Blister growth in zirconium alloys: experimentation Application of the surface-mobility stress corrosion
and modeling, G. Domizzi, R.A. Enrique, J. Ovejero- cracking mechanism to nuclear materials, J.R.
Garcia and G.C. Buscaglia.. 1.1 1. ee ee ees 36 ED 6b ahaa woe wed DEO 0 3 eK eee ES 139
Hydriding kinetics of Zircaloy-4 in hydrogen gas, G.
Meyer, M. Kobrinsky, J.P. Abriata and J.C. Bolcich . 48 Section 4. Nuclear power performance
; , Simulation of isothermal fission gas release, A. Denis
1.4. Mechanical behavior ra Se - 5 0:4 & 4.0 400-4 0 8-0-0 ba 8-8 149
Measurement and prediction of texture development BACO (BArra COmbustible) code version 2.20: a
during a rolling sequence of Zircaloy-4 tubes, R.A. thermo-mechanical description of a nuclear fuel
Lebensohn, M.I. Gonzalez, C.N. Tomé and A.A. rod, A.C. Marino, E.J. Savino and S. Harriague ... 155
I cteae a eg 57 Irradiation of Argentine (U,PuJO, MOX fuels. Post-
Intergranular thermal stresses in zirconium — effects irradiation results and experimental analysis with
on X-rays macrostress measurements, M. Ortiz and the BACO code, A.C. Marino, E. Pérez and P.
SL EE 4-06.66 5-4 0 0 2 é-26 0 0 Owe es 65 Adelfang .. cc escesccreseesceesens 169
1.5. Radiation damage Section 5. Nuclear waste immobilization
235 Immobilization of simulated high-level liquid wastes in
Irradiation growth kinetics in “~~U -doped zirconium,
G.D.H. Coccoz. A.M. Fortis and H.C. Gonzélez 73 sintered borosilicate, aluminosilicate and alumi-
_ noborosilicate glasses, A.M. Bevilacqua, N.B.
1.6. Surface oxidation and corrosion Messi de Bernasconi, D.O . Russo, M.A. Audero, M.E.
pe ee ee ee ee ee 187
Growth and characterization of oxide layers on zirco-
nium alloys, A.J.G. Maroto, R. Bordoni, M. Villegas, ee fs ok 56 re ot Se OE ee 195
A.M. Olmedo, M.A. Blesa, A. Iglesias and P. Koenig . eeeee ee en ee a