ANS / Publications / Journals / Nuclear Technology

Effects of Different Types of Void Volumes on the Radial Temperature Distribution of Fuel Pins

H. Kämpf, G. Karsten

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 288-300

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28783

Mechanical and Thermal Analysis of Cylindrical Fuel Elements During Off-Normal Conditions After Extended Burnup

T. R. Bump

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 301-308

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28784

Evaluation of a Model for Predicting Fast-Reactor Fuel-Pin Deformations

K. R. Merckx

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 309-316

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28785

Performance Analysis of a Mixed-Oxide LMFBR Fuel Pin

C. M. Cox, F. J. Homan

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 317-325

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28786

Fast Reactor Fuel Performance Model Development

A. Boltax, P. Murray, A. Biancheria

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 326-337

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28787

Non-Steady-State Factors in Models for Swelling of Oxide Fuels

D. P. Hines, S. Oldberg, E. L. Zebroski

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 338-345

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28788

The Growth and Stability of Voids in Irradiated Metals

R. Bullough, B. L. Eyre, R. C. Perrin

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 346-355

Fuel Element Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28789

Optimization of Fuel Loadings for High Power Test Reactors

H. J. Reilly, L. E. Peters, Jr.

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 356-363

Fuel Cycle / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28790

An Estimate of the Enhancement of Fission Product Release from Molten Fuel by Thermally Induced Internal Circulation

M. H. Fontana

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 364-375

Fuel / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28791

A Long Range Planning Model of the USAEC Gaseous Diffusion Plant

Henry Stone, A. De La Garza, R. L. Hoglund

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 376-395

Economic / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28792

Cryogenic Tensile Properties of Irradiated Beryllium, Titanium, and Aluminum Alloys

J. R. Coombe, R. P. Shogan

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 396-401

Material / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28793

Testing for Incipient Failure of Relays in Reactor Circuits

John Perreault, Lawrence Ruby

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 402-407

Instrument / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28794

A New Boron Analysis Method

J. Weitman, N. Dåverhög, S. Farvolden

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 408-415

Analysis / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28795

Remote Analyses by Atomic Absorption Spectrophotometry

W. R. Sovereign, E. R. Ebersole, R. Villarreal, W. A. Hareland

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 416-421

Technique / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28796

Hydraulic Impedance: A Tool for Predicting Boiling Loop Stability

T. T. Anderson

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 422-433

Technique / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28797

Effective Alpha Activity and Self-Absorption Alpha Range in ²³⁸PuO₂ Microspheres

Gary N. Huffman, Carl J. Kershner

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Pages 434-438

Radioisotope / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28798

Gamma-Ray Buildup Factor Coefficients for Concrete and other Materials

D. K. Trubey

Nuclear Technology / Volume 9 / Number 3 / September 1970 / Page 439

Shielding / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28799