Beschreibung
InhaltsangabeOpening Session.- Opening Address.- Keynote Speech.- Neutron Data of Transactinium Isotopes (Fission Reactor Core Constituents Mainly).- Invited Papers.- 238U, Issues Resolved and Unresolved.- Besoins en Donnees Nucleaires pour les Reacteurs a Neutrons Thermiques.- Contributed Papers.- Modified Scattering Matrices to Improve Transport Calculations with Approximate Fission-Source Matrices.- Comparisons of Calculated Self Shielding Factors with Measured Values for 239Pu, 235U, Fe and Na.- Fast Neutron Induced Fission Cross Section for Pu-239.- On the Neutron Inelastic-Scattering Cross Sections of 232Th, 233U, 235U, 238U, 239Pu and 240Pu.- Recent Results of Neutron Inelastic Scattering to Higher-Excited Levels in 232Th and 238U.- Mesures Absolues de 240Pu(n,f), 242Pu(n,f) ET 237Np(n,f) A L'energie Incidente de 2,5 MeV.- Absolute Measurements of 235U and 239Pu Fission Cross Section Induced by 14.7 MeV Neutrons.- Measurements of the 14 MeV Fission Cross Sections for 235U and 239Pu.- Study of Resonance Neutron Cross Section Structure of U-238 and Pu-239.- Neutron Total Cross Section Measurements of 238U at the Fe-Filtered Neutron Energy Bands in keV Region.- The 24lPu(n,f) Cross-Section and its Normalization.- Neutron Induced Fission Cross Section of 244Pu.- Mesure De (Mathtype) et (Mathtype) pour la fission de 235th, 235u et 237Np Induite Par Neutrons D'Energie Comprise Entre 1 et 15 MeV.- Neutron Data of Structural Materials, Coolants and Shielding of Fission Reactors.- Invited Paper.- Convergence of Integral and Differential Cross-Section Data for Structural Materials.- Contributed Papers.- The Integral Check of Neutron Cross Section Data for Reactor Structural Materials by Measurement and Analysis of Neutron Spectra.- Integral Capture Cross Section Measurements of Some Structural Materials in a Fast Spectrum.- Etude Experimentale de la Capture des Materiaux de Structure Dans des Reseaux a Neutrons Rapides par la Methode K? = 1.- Ajustement des Donnees Nucleaires de Materiaux de Structure a Partir d'Experiences Integrales Specifiques.- Adjustment of Neutron Multigroup Cross Sections with Error Covariance Matrices to Deep Penetration Integral Experiments.- 54Fe Neutron Capture Cross Section.- High Resolution Neutron Capture Cross Section Measurements of 56Fe.- Neutron Resonance Structure of 54Fe and 56Fe From High Resolution Total Cross Section Experiments.- Resonance Parameters of 57Fe.- Total Cross Section Measurements of Thermal and 24 keV Neutrons for Crystalline Materials.- Search for Gas Producing Reactions in Thermal Reactors.- Study of the 22Na(n,p)22Ne Reaction in the Neutron Energy Range up to 1 keV.- Gamma-Rays from Capture of 400-keV Neutrons.- Measurement of (n,?) Cross Sections for Cr, Fe and Ni at 14 MeV Neutron Energy.- Fast Neutron Interaction with Chromium-52 and Molybdenum-92.- Nuclear Data Pertaining to Fission Reactor Fuel Cycles and Fission Products.- Invited Paper.- Nuclear Data Needs for Uranium-Plutonium Fuel Cycle Development.- Contributed Papers.- Donnees de Base Dans le Cycle du Combustible des Reacteurs a Neutrons Rapides Bilan et Perspective.- The Taco Experiment for the Determination of Integral Neutron Cross-Sections in a Fast Reactor.- Integral Experiments to Measure the Production Rates of 242Cm and 244Cm in Fast Reactor Spectra.- Determination Experimentale des Sections Efficaces des Isotopes de Pu, Am, Cm Dans un Spectre de Neutrons de Reacteur A EAU.- Reactor Irradiations of 242Pu, Comparison of Measured and Calculated Yields of 244Pu, 243Am and 244Cm, and Study of the Fission Product Yields.- Experimental Validation of Irradiated Fuel Inventories Calculated by the Fispin Code Fuels.- A Critical Review of Resonance Integrals and Postirradiation Fuel Analysis for Important Isotopes of Am and Cm.- Analysis of Neutron Cross Sections for the Formation of Pu-236 and Co-58,60 in both Thermal and Fast Reactors.- Neutron Induced Fission Cross Section of 238Pu in the Energy Range From 5 eV to 10 MeV.- N
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