ANS / Publications / Journals / Nuclear Technology

Theories of Swelling and Gas Retention in Ceramic Fuels

Brian R. T. Frost

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 128-140

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28803

A Fission Gas Swelling Model Incorporating Re-Solution Effects

C. C. Dollins, H. Ocken

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 141-147

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28804

A Statistical Fuel Swelling and Fission Gas Release Model

H. R. Warner, F. A. Nichols

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 148-166

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28805

Interpretations of Fission Gas Behavior in Refractory Fuels

R. L. Ritzman, A. J. Markworth, W. Oldfield, W. Chubb

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 167-187

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28806

Some Considerations of the Behavior of Fission Gas Bubbles in Mixed-Oxide Fuels

Che-Yu Li, S. R. Pati, R. B. Poeppel, R. O. Scattergood, R. W. Weeks

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 188-194

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28807

A Computer Program to Predict the Performance of UO₂ Fuel Elements Irradiated at High Power Outputs to a Burnup of 10 000 MWd/MTu

M. J. F. Notley

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 195-204

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28808

Comethe II-A Computer Code for Predicting the Mechanical and Thermal Behavior of a Fuel Pin

R. Godesar, M. Guyette, N. Hoppe

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 205-217

Fuel Performance Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28809

The Nuclear Criticality Safety Aspects of Plutonium-238

Richard A. Wolfe

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 218-228

Reactor Siting / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28810

Accurate Absolute Determination of Fission Densities in Fuel Rods by Means of Solid-State Track Detectors

M. De Coster, D. Langela

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 229-232

Fuel / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28811

Internal Gas Pressure Behavior in Mixed-Oxide Fuel Rods Fuels During Irradiation

T. B. Burley, M. D. Freshley

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 233-241

Fuel / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28812

Utility Incentives for Implementing Crossed-Progeny Fueling

L. W. Lang

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 242-249

Economic / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28813

Thermal Convection Loop Tests of Nb-1% Zr Alloy in Lithium at 1200 and 1300°C

C. E. Sessions, J. H. DeVan

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 250-259

Material / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28814

Environmentally Aggravated Fatigue Cracking of Zircaloy-2

Lee A. James

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 260-267

Material / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28815

Measurement of Low Levels of Iodine-131 in Reactor Atmospheres

V. C. Furtado, T. J. Kneip, M. Eisenbud

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 268-273

Technique / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28816

Nuclear Engineering Internships Education

Robert L. Carter

Nuclear Technology / Volume 9 / Number 2 / August 1970 / Pages 274-277

Education / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28817