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Volume 59

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Neutron Transport with Temperature Feedback

A. Belleni-Morante

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 56-58

January 1976

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Multigroup Representations of Slowing-Down Kernels in H2O

Milton C. Edlund, Richard B. Jones, P. F. Zweifel

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 58-60

January 1976

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On the Generalized Perturbation Methods in Time-Dependent Problems

A. Gandini

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 60-63

January 1976

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A Finite Element Synthesis Method

Shi-tien Yang, A. F. Henry

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 63-67

January 1976

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Comments on “Analysis of the Microfission Reactor Concept”

F. Winterberg

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 68-70

January 1976

Reply to “Comments on ‘Analysis of the Microfission Reactor Concept’”

R. K. Cole, Jr., J. H. Renken

Nuclear Science and Engineering / Volume 59 / Number 1 / Page 70

January 1976

From Scientific Search to Atomic Industry

Karl Cohen

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 71-72

January 1976

Nuclear-Reactor Analysis

G. C. Pomraning

Nuclear Science and Engineering / Volume 59 / Number 1 / Pages 72-73

January 1976

Table of Contents

Nuclear Science and Engineering / Volume 59 / Number 3

March 1976

Monte Carlo Surface Density Solution to the Dirichlet Heat Transfer Problem

T. J. Hoffman, N. E. Banks

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 205-214

March 1976

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A General Method for Generating Effective Resonance Cross Sections for Heterogeneous Media

K. D. Kirby, R. A. Karam

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 215-230

March 1976

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The 165Ho(n, γ) Standard Cross Section from 3 to 450 keV

R. L. Macklin

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 231-236

March 1976

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Spectra of Bremsstrahlung Produced in Very Thick Lead Targets by 15-, 20-, and 25-MeV Electrons

Takashi Nakamura, Hideo Hirayama

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 237-245

March 1976

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Moments Method Calculation of the Energy-Dependent Neutron Flux Due to a Point-lsotropic Fission Source in an Infinite Medium of Sodium

E. E. Morris

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 246-260

March 1976

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Effect of Anisotropy in Scattering on Neutron Angular Flux Inside Slabs

Lakshmi Rangaswamy, L. S. Kothari, Feroz Ahmed

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 261-270

March 1976

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Accuracy of the Quasistatic Method for Two-Dimensional Thermal Reactor Transients with Feedback

H. L. Dodds, Jr.

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 271-276

March 1976

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Incorporation of Spectral Effects into One-Group Nodal Simulators

Martin Becker

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 276-278

March 1976

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Anisotropic Neutron Transport Without Legendre Expansions

J. P. Odom, J. K. Shultis

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 278-281

March 1976

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Remarks on “Neutron Transport with Temperature Feedback”

Aldo Belleni-Morante

Nuclear Science and Engineering / Volume 59 / Number 3 / Page 282

March 1976

Techniques in Nuclear Structure Physics

T. F. Parkinson

Nuclear Science and Engineering / Volume 59 / Number 3 / Pages 283-284

March 1976

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