ANS / Publications / Journals / Nuclear Technology

Application of RETRAN-02 to Plant Transient Analysis and Issue Resolution

Pedro Mata Alonso, Rafael Otero Carrasco, Pablo Garcia Sedano

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 7-15

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34513

Developing RETRAN Input for Redundant Networks

Kenneth J. Doran, Ronald E. Engel, Randall S. May

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 16-21

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34514

Steam Line Break Analysis Comparison to the Wolf Creek Updated Safety Analysis Report

Glenn J. Neises, Terry J. Garrett

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 22-35

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34515

Core Boiling During Midloop Operation

Nobuyuki Fujita, David A. Rice

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 36-46

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34516

A Survey of the Oyster Creek Reload Licensing Model

Mansur A. Alammar

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 47-52

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34517

Palo Verde Nuclear Generating Station Midloop Condition RETRAN Model

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 53-64

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34518

Response Surface Development Using RETRAN

Ronald E. Engel, John M. Sorensen, Randall S. May, Kenneth J. DOran, N. G. Trikouros, Eugene S. Mozias

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 65-81

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34519

Analysis of a Neutron Flux Oscillation Event at LaSalle-2

Fumimasa Araya, Kazuo Yoshida, Masashi Hirano, Yukihisa Yabushita

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 82-91

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34520

Consistent Initialization of Pressure and Flow Parameters for BWR Transient Analysis at Off-Rated Power and Flow Conditions

Bennett J. Gitnick

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 92-104

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34521

The RETRAN-03 Computer Code

M. P. Paulsen, J. H. McFadden, C. E. Peterson, J. A. McClure, G. C. Gose, P. J. Jensen

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 105-115

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34522

Compiler Issues Associated with Safety-Related Software

Lynn R. Feinauer

Nuclear Technology / Volume 93 / Number 1 / January 1991 / Pages 116-122

Technical Note / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34523

Sensitivity Analysis of Thermal-Hydraulic Parameters and Probability Estimation of Boiling Transition in a Standard BWR/6

Smaeil M. Aceil, Doyle R. Edwards

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 123-130

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34498

Sequential Probability Ratio Test for Nuclear Plant Component Surveillance

Kenny C. Gross, Keith E. Humenik

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 131-137

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34499

Core Simulations Using Actual Detector Readings for a Canada Deuterium Uranium Reactor

Bong Ghi Kim, In Sup Kim, Seong Yun Kim

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 138-146

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34500

Decontamination of Liquid-Metal Fast Breeder Reactor Components for Reuse: The French Experience

Patrick Michaille, Jean-Claude Moroni, Irma Lambert

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 147-157

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34501

The Potential for Catastrophic Fuel Damage During Beyond-Design-Basis Overpressurization Accidents in a Boiling Water Reactor

David J. Diamond, Dragan Mirkovic, Chia-Jung Hsu, Robert Fitzpatrick

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 158-165

Technical Paper / Nuclear Reactor Safet / dx.doi.org/10.13182/NT91-A34502

Countercurrent Flow Limitation Model for RELAP5/MOD3

Richard A. Riemke

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 166-173

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34503

Small-Scale Experiments Conducted to Support Investigation of HypotheticalDesign-Basis LOCA Phenomena in the Advanced Test Reactor

Thomas K. Larson, Keith G. Condie, Steven T. Polkinghorne, Brent J. Buescher

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 174-194

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34504

A Statistical Model for Prediction of Fuel Element Failure Using the Markov Process and Entropy Minimax Principles

Kie-Yong Choi, Young-Ku Yoon, Soon-Heung Chang

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 195-205

Technical Paper / Nuclear Fuel / dx.doi.org/10.13182/NT91-A34505

Development of Partitioning and Transmutation Technology for Long-Lived Nuclides

Tadashi Inoue, Masahiro Sakata, Hajime Miyashiro, Tetsuo Matsumura, Akihiro Sasahara, Nobuya Yoshiki

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 206-220

Technical Paper / Nuclear Fuel Cycle / dx.doi.org/10.13182/NT91-A34506

Severe Accident Zircaloy Oxidation/Hydrogen Generation Behavior Noted from In-Pile Test Data

August W. Cronenberg

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 221-239

Technical Paper / Material/Nuclear Safety / dx.doi.org/10.13182/NT91-A34507

RELAP5 Simulation of Advanced Test Reactor Scaled LOCA Experiments 7, 8, 12, and 14

Steven T. Polkinghorne, Thomas K. Larson, Brent J. Buescher

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 240-251

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT91-A34508

Development of a Computer Code to Simulate Aerosol Behavior in a Vented Containment

A. K. Ahluwalia, A. M. C. Chan

Nuclear Technology / Volume 93 / Number 2 / February 1991 / Pages 252-259

Technical Note / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34509

A High-Breeding Fast Reactor with Fission Product Gas Purge/Tube-in-Shell Metallic Fuel Assemblies

Toru Hiraoka, Kiyoshi Sako, Hideki Takano, Takeshi Ishii, Mitsuru Sato

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 305-329

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34526

Reactor Design of the Advanced Neutron Source

John M. Ryskamp, Douglas L. Selby, R. Trenton Primm III

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 330-349

Technical Paper / Fission Reactor / dx.doi.org/10.13182/NT91-A34527

Control Room Habitability During Severe Accidents

Raymond P. Siu

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 350-356

Technical Paper / Nuclear Reactor Safety / dx.doi.org/10.13182/NT91-A34528

Studies of Neutron Shielding Properties of Ilmenite-Magnetite Concrete Using A ²⁵²Cf Source

S. I. Bhuiyan, F. U. Ahmed, A. S. Mollah, M. A. Rahman

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 357-361

Technical Paper / Nuclear Fuel Cycle / dx.doi.org/10.13182/NT91-A34529

Numerical Study of Hydrodynamics in Pulsed Solvent Extraction Columns Using the Dynac Code

Masahiro Nabeshima

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 362-374

Technical Paper / Enrichment and Reprocessing System / dx.doi.org/10.13182/NT91-A34530

Purex Processing of 10-Year Light Water Reactor Fuel: Flow Sheet

Okan H. Zabunoglu, Bernard I. Spinrad

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 376-381

Technical Paper / Enrichment and Reprocessing System / dx.doi.org/10.13182/NT91-A34531

A Method to Evaluate the Maximum Allowable Temperature of Spent Fuel in Dry Storage During a Postulated Accident

Masami Mayuzumi, Takeo Onchi

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 382-388

Technical Paper / Radioactive Waste Management / dx.doi.org/10.13182/NT91-A34532

Flux and Fluence Determination Using the Material Scrapings Approach

Michael P. Manahan, Sr., Hassan S. Basha

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 389-398

Technical Paper / Material / dx.doi.org/10.13182/NT91-A34533

High-Temperature Corrosion Testing of Alumina and Zirconia in Uranium Hexafluoride Environment

Simon C. P. Wang, Clayton Collins, Samim Anghaie, E. Dow Whitney

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 399-411

Technical Paper / Material / dx.doi.org/10.13182/NT91-A34534

A Step-Heating Dilatometry Method to Measure the Change in Length Due to Annealing of a SiC Temperature Monitor

Toyohiko Yano, Kazunari Sasaki, Tadashi Maruyama, Takayoshi Iseki, Masahiko Ito, Shoji Onose

Nuclear Technology / Volume 93 / Number 3 / March 1991 / Pages 412-415

Technical Note / Material / dx.doi.org/10.13182/NT91-A34535