Volume 56
Number 4
- Table of Contents
- Comments on “Measurements of Anisotropic Neutron Diffusion Coefficients in Square Lattices of Aluminum in Light Water by the Pulsed Neutron Method”
- Comments on “On the Randomness of a Neutron’s Kinetic Energy as It Slows Down by Elastic Collisions in an Infinite Medium”
- Response to “Comments on ‘On the Randomness of a Neutron’s Kinetic Energy as It Slows Down by Elastic Collisions in an Infinite Medium’”
- Annual Review of Nuclear Science, Volume 23
A Generalized Diffusion Theory for Calculating the Neutron Transport Scalar Flux
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 321-339
Technical Paper / dx.doi.org/10.13182/NSE75-A26680
Variational Functionals Which Admit Discontinuous Trial Functions
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 340-353
Technical Paper / dx.doi.org/10.13182/NSE75-A26681
Nonlinear Oscillations in a Reactor with Two Temperature Coefficients
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 354-359
Technical Paper / dx.doi.org/10.13182/NSE75-A26682
Calculational Methods for Nuclear Heating-Part I: Theoretical and Computational Algorithms
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 360-380
Technical Paper / dx.doi.org/10.13182/NSE75-A26683
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 381-398
Technical Paper / dx.doi.org/10.13182/NSE75-A26684
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 399-410
Technical Paper / dx.doi.org/10.13182/NSE75-A26685
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 411-422
Technical Note / dx.doi.org/10.13182/NSE75-A26686
Zero-Power Noise Analysis in a Reactor with Two Weakly Coupled Asymmetrical Fission Zones
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 422-427
Technical Note / dx.doi.org/10.13182/NSE75-A26687
A Benchmark Experiment of Neutron Propagation in Iron Used to Test ENDF/B Cross-Section Data
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 427-430
Technical Note / dx.doi.org/10.13182/NSE75-A26688
Power Reactor Calculations with the Finite Element Program FEM 2D
Nuclear Science and Engineering / Volume 56 / Number 4 / April 1975 / Pages 431-433
Technical Note / dx.doi.org/10.13182/NSE75-A26689
