Volume 9
Number 1 | Number 2 | Number 3 | Number 4 | Number 5 | Number 6
Number 1
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 10-23
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28723
An Analysis of Fast Neutron Effects on Void Formation And Creep in Metals
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 24-30
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28724
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 31-39
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28725
Theoretical Analysis of Cladding Stresses and Strains Produced by Expansion of Cracked Fuel Pellets
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 40-46
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28726
Axial Ratchetiing of Fuel Under Pressure Cycling Conditions
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 47-59
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28727
Crash-A Computer Program for the Evaluation of the Creep and Plastic Behavior of Fuel-Pin Sheaths
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 60-69
Fuel Cladding Model / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28728
A Three-Dimensional Method for Design Studies of Xenon-Induced Spatial Power Oscillations
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 70-82
Reactor / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28729
The Nuclear Performance of Fusion Reactor Blankets
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 83-92
Reactor / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28730
Calculational Models for Fast Reactor Fuel-Cycle Analysis
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 93-106
Fuel / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28731
The Effects of Contaminants in Methane as a Proportional Tube Counting Gas
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 107-111
Instrument / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28732
Evaluation of Cdte as an Integral Gamma-Ray Counter
Nuclear Technology / Volume 9 / Number 1 / July 1970 / Pages 112-119
Instrument / Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material / dx.doi.org/10.13182/NT70-A28733
Number 2
Theories of Swelling and Gas Retention in Ceramic Fuels
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
Number 3
Effects of Different Types of Void Volumes on the Radial Temperature Distribution of Fuel Pins
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
Number 4
Neutron-Energy Spectra for Fast Reactor Irradiation Effects
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 454-507
Department / Reactor / dx.doi.org/10.13182/NT70-A28760
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 508-515
Reactor / dx.doi.org/10.13182/NT70-A28761
Plutonium Recycle Studies for the Sena Pwr Reactor
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 516-527
Fuel Cycle / dx.doi.org/10.13182/NT70-A28762
Burst Strength of EBR-II Irradiated Fuel Pin Sections
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 528-535
Uranium-Plutonium Mixed Oxide Sol-Gel Irradiation Experiments
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 536-549
Effect of Irradiation on the Elevated Temperature Fracture of Selected Face-Centered Cubic Alloys
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 550-560
Material / dx.doi.org/10.13182/NT70-A28765
Incoloy 800: Enhanced Resistance to Radiation Damage
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 561-571
Material / dx.doi.org/10.13182/NT70-A28766
Reentry Protection for Radioisotope Heat Sources
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 572-583
Radioisotope / dx.doi.org/10.13182/NT70-A28767
An Analog Computer Controlled Gamma-Ray Spectrometer for Comparative Activation Analysis
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 584-590
Radioisotopes / dx.doi.org/10.13182/NT70-A28768
Analysis of Gamma-Ray Spectroscopy Data
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 591-604
Analysis / dx.doi.org/10.13182/NT70-A28769
Calculation of the Long-Lived Induced Activity in Soil Produced by 200-MeV Protons
Nuclear Technology / Volume 9 / Number 4 / October 1970 / Pages 605-614
Analysis / dx.doi.org/10.13182/NT70-A28770
Number 5
Physics Of Operating Boiling Water Reactors
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 622-633
Paper / Reactor / dx.doi.org/10.13182/NT70-A28736
Physics of Operating Pressurized Water Reactors
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 634-639
Paper / Reactor / dx.doi.org/10.13182/NT70-A28737
Physics Performance of Shippingport
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 640-650
Paper / Reactor / dx.doi.org/10.13182/NT70-A28738
Experience with Xenon Oscillations
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 651-661
Paper / Reactor / dx.doi.org/10.13182/NT70-A28739
Average Thermal Neutron Capture Cross Sections of 198Au, 65Ni, And 66Cu
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 662-666
Paper / Reactor / dx.doi.org/10.13182/NT70-A28740
Spontaneous Deposition of Polonium-210 from Chloride Solution
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 667-672
Paper / Chemical Processing / dx.doi.org/10.13182/NT70-A28741
Achieving High Exposure in Metallic Uranium Fuel Elements
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 673-681
Paper / Fuel / dx.doi.org/10.13182/NT70-A28742
Single- and Two-Phase Pressure Drops on a 16-Rod Bundle
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 682-693
Paper / Fuel / dx.doi.org/10.13182/NT70-A28743
The Application of Reliability Margin Analysis to Fuel Element Performance
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 694-699
Paper / Fuel / dx.doi.org/10.13182/NT70-A28744
Helium Production in EBR-II Irradiated Stainless Steel
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 700-710
Paper / Fuel / dx.doi.org/10.13182/NT70-A28745
Rate Controlling Factors in the Carbothermic Synthesis of Advanced Fuels
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 711-715
Flowing Sodium Capsules in the GETR
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 716-721
Paper / Material / dx.doi.org/10.13182/NT70-A28747
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 722-735
Paper / Aerospace / dx.doi.org/10.13182/NT70-A28748
A Neutron Detection System for Operation in Very High Gamma Fields
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 736-743
Paper / Instrument / dx.doi.org/10.13182/NT70-A28749
Evaluation Of A Neutron Detection System In A Cobalt-6O Field Of 107 R/H
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 744-745
Paper / Instrument / dx.doi.org/10.13182/NT70-A28750
Techniques for Two-Dimensional Gamma-Ray Scanning of Reactor Fuel Element Sections
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 746-754
Paper / Technique / dx.doi.org/10.13182/NT70-A28751
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 755-761
Paper / Technique / dx.doi.org/10.13182/NT70-A28752
Estimation Techniques for Far-field Exposure Contributions
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 762-766
Paper / Technique / dx.doi.org/10.13182/NT70-A28753
Prediction of the Incipient Boiling Conditions Following a Blocked Lmfbr Subassembly Accident
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 767-770
Note / Reactor Siting / dx.doi.org/10.13182/NT70-A28754
Ductility Loss in Fast Reactor Irradiated Stainless Steel
Nuclear Technology / Volume 9 / Number 5 / November 1970 / Pages 771-772
Note / Material / dx.doi.org/10.13182/NT70-A28755
Number 6
A Bootstrap Concept of a Safety Test Facility
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 780-785
Reactor Siting / dx.doi.org/10.13182/NT70-A28709
Civil Defense Implications of a Pressurized Water Reactor in a Thermonuclear Target Area
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 786-795
Reactor / dx.doi.org/10.13182/NT70-A28710
Evaluation of Isocheck and Invent Fuel Inventory Calculational Models
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 796-806
Fuels Cycle / dx.doi.org/10.13182/NT70-A28711
Uranium-233-Bearing Salt Preparation for the Molten Salt Reactor Experiment
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 807-813
Chemical Processing / dx.doi.org/10.13182/NT70-A28712
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 814-820
Chemical Processing / dx.doi.org/10.13182/NT70-A28713
The Feasibility of Incorporating Radioactive Wastes in Asphalt or Polyethylene
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 821-829
Radioactive Waste / dx.doi.org/10.13182/NT70-A28714
The Synthetic Actinides-From Discovery to Manufacture
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 830-850
Radioisotope / dx.doi.org/10.13182/NT70-A28715
Pulsed-Neutron Activation Analysis System for Short-Lived Radioisotopes
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 851-855
Analysis / dx.doi.org/10.13182/NT70-A28716
Proton Microprobe Analysis of the Surface opf Stranded Wire in the Lunar Module
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 856-860
Analysis / dx.doi.org/10.13182/NT70-A28717
Laboratory and Environmental Mineral Analysis using a Californium-252 Neutron Source
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 861-874
Analysis / dx.doi.org/10.13182/NT70-A28718
A 244Cm-Be Isotopic Neutron Source
Nuclear Technology / Volume 9 / Number 6 / December 1970 / Pages 875-878
Technique / dx.doi.org/10.13182/NT70-A28719
