Home / Publications / Journals / Nuclear Technology / Volume 76 / Number 2
Determination of Isotopic Ratios from Fuel Burnup
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 203-208
Technical Paper / Fuel Cycle / dx.doi.org/10.13182/NT87-A33874
ELOCA-A: A Code for Radial and Axial Behavior of CANDU Fuel Elements at High Temperatures
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 209-220
Technical Paper / Nuclear Fuel / dx.doi.org/10.13182/NT87-A33875
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 221-228
Technical Paper / Radioactive Waste Management / dx.doi.org/10.13182/NT87-A33876
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 229-240
Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT87-A33877
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 241-247
Technical Paper / Analyse / dx.doi.org/10.13182/NT87-A33878
Simulation of the Transient Response of Ionization Chambers to Bias Voltage Perturbations
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 248-259
Technical Paper / Technique / dx.doi.org/10.13182/NT87-A33879
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 263-275
Performance of Borosilicate Glass High-Level Waste Forms in Disposal Systems / Radioactive Waste Management / dx.doi.org/10.13182/NT87-A33880
A Guillotine Tube Rupture Modeling Technique Using RETRAN-02
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 279-289
Fourth International Retran Meeting / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT87-A33881
A Parametric Study of an Anticipated Transient Without Scram in a Westinghouse Four-Loop Plant
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 290-302
Fourth International Retran Meeting / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT87-A33882
Moderator Feedback Effects in Two-Dimensional Nodal Methods for Pressurized Water Reactor Analysis
Nuclear Technology / Volume 76 / Number 2 / February 1987 / Pages 303-307
Technical Note / Fission Reactor / dx.doi.org/10.13182/NT87-A33883