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Volume 130

Number 1

A Pin Power Reconstruction Method for CANDU Reactor Cores Based on Coarse-Mesh Finite Difference Calculations

Hyung-Seok Lee, Won Sik Yang, Man Gyun Na, Hangbok Choi

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 1-8

Technical Paper / Fission Reactors / dx.doi.org/10.13182/NT00-A3072

Fission Gas Release from UO2+x in Defective Fuel Rods

Yeon Soo Kim

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 9-17

Technical Paper / Fuel Cycle and Management / dx.doi.org/10.13182/NT00-A3073

RELAP5 Simulation of Thermal-Hydraulic Behavior in a CANDU Reactor - Assessments of RD-14 Experiments

Sukho Lee, In-Goo Kim

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 18-26

Technical Paper / Thermal Hydraulics / dx.doi.org/10.13182/NT00-A3074

Development of a Dynamic Event Tree for a Pressurized Water Reactor Steam Generator Tube Rupture Event

Chun-Chang Chao, Chin-Jang Chang

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 27-38

Technical Paper / Thermal Hydraulics / dx.doi.org/10.13182/NT00-A3075

Heat Transfer Modeling During a Vapor Explosion

Georges Berthoud

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 39-58

Technical Paper / Thermal Hydraulics / dx.doi.org/10.13182/NT00-A3076

Hydrogen Water Chemistry Technology in Boiling Water Reactors

Chien C. Lin

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 59-70

Technical Paper / Nuclear Plant Operations and Control / dx.doi.org/10.13182/NT00-A3077

Analysis of Radionuclide Release from Spent Ion-Exchange Resins

Sol-Il Su, Man-Sung Yim

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 71-88

Technical Paper / Radioactive Waste Management and Disposal / dx.doi.org/10.13182/NT00-A3078

Alternatives to High-Level Waste Vitrification: The Need for Common Sense

Jimmy T. Bell

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 89-98

Technical Paper / Radioactive Waste Management and Disposal / dx.doi.org/10.13182/NT00-A3079

A Comparison of Natural Convection Heat Transfer for a Staggered Versus an Aligned Array of Horizontal Spent Nuclear Fuel Rods Within a Rectangular Enclosure

C. Eric Triplett, Robert E. Canaan, Dale E. Klein

Nuclear Technology / Volume 130 / Number 1 / April 2000 / Pages 99-110

Technical Note / Thermal Hydraulics / dx.doi.org/10.13182/NT00-A3080