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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Reboot: Nuclear needs a success . . . anywhere
The media have gleefully resurrected the language of a past nuclear renaissance. Beyond the hype and PR, many people in the nuclear community are taking a more measured view of conditions that could lead to new construction: data center demand, the proliferation of new reactor designs and start-ups, and the sudden ascendance of nuclear energy as the power source everyone wants—or wants to talk about.
Once built, large nuclear reactors can provide clean power for at least 80 years—outlasting 10 to 20 presidential administrations. Smaller reactors can provide heat and power outputs tailored to an end user’s needs. With all the new attention, are we any closer to getting past persistent supply chain and workforce issues and building these new plants? And what will the election of Donald Trump to a second term as president mean for nuclear?
As usual, there are more questions than answers, and most come down to money. Several developers are engaging with the Nuclear Regulatory Commission or have already applied for a license, certification, or permit. But designs without paying customers won’t get built. So where are the customers, and what will it take for them to commit?
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The Effects of Fast Flux Irradiation on the Mechanical Properties and Dimensional Stability of Stainless Steel
T. T. Claudson, R. W. Barker, R. L. Fish
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 | doi.org/10.13182/NT70-A28723
An Analysis of Fast Neutron Effects on Void Formation And Creep in Metals
S.D. Harkness, J. A. Tesk, Che-Yu Li
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 | doi.org/10.13182/NT70-A28724
New Correlations Involving the Low-Cycle Fatigue and Short-Term Tensile Behavior of Irradiated and Unirradiated 304 and 316 Stainless Steel
J. B. Conway, J. T. Berling, R. H. Stentz
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 | doi.org/10.13182/NT70-A28725
Theoretical Analysis of Cladding Stresses and Strains Produced by Expansion of Cracked Fuel Pellets
J. H. Gittus, D. A. Howl, H. Hughes
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 | doi.org/10.13182/NT70-A28726
Axial Ratchetiing of Fuel Under Pressure Cycling Conditions
Eliot Duncombe, Ivan Goldberg
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 | doi.org/10.13182/NT70-A28727
Crash-A Computer Program for the Evaluation of the Creep and Plastic Behavior of Fuel-Pin Sheaths
M. Guyette
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 | doi.org/10.13182/NT70-A28728
A Three-Dimensional Method for Design Studies of Xenon-Induced Spatial Power Oscillations
R. C. Kern, W. C. Coppersmith, Z. R. Rosztoczy
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 | doi.org/10.13182/NT70-A28729
The Nuclear Performance of Fusion Reactor Blankets
D. Steiner
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 | doi.org/10.13182/NT70-A28730
Calculational Models for Fast Reactor Fuel-Cycle Analysis
Thomas J. Hirons, R. Douglas O'Dell
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 | doi.org/10.13182/NT70-A28731
The Effects of Contaminants in Methane as a Proportional Tube Counting Gas
F. E. Armstrong, W. D. Howell, D. W. Whitlock
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 | doi.org/10.13182/NT70-A28732
Evaluation of Cdte as an Integral Gamma-Ray Counter
H. H. Nichols
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 | doi.org/10.13182/NT70-A28733
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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | doi.org/10.13182/NT70-A28817
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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | 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 | doi.org/10.13182/NT70-A28799
Neutron-Energy Spectra for Fast Reactor Irradiation Effects
D. Okrent, W. B. Loewenstein, A. D. Rossin, A. B. Smith, B. A. Zolotar, J. M. Kallfelz
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 454-507
Department | Reactor | doi.org/10.13182/NT70-A28760
In-Place Testing of the Hanford Reactor Charcoal Confinement Filter Systems using Iodine Tagged with Iodine-131
J. E. Mecca, J. D. Ludwick
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 508-515
Reactor | doi.org/10.13182/NT70-A28761
Plutonium Recycle Studies for the Sena Pwr Reactor
J. Debrue, P. Deramaix, F. De Waegh
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 516-527
Fuel Cycle | doi.org/10.13182/NT70-A28762
Burst Strength of EBR-II Irradiated Fuel Pin Sections
R. L. Fish, J. J. Holmes, R. D. Leggett
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 528-535
Fuel | doi.org/10.13182/NT70-A28763
Uranium-Plutonium Mixed Oxide Sol-Gel Irradiation Experiments
C. Lepscky, P. L. Rotoloni, G. Testa, G. Trezza
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 536-549
Fuel | doi.org/10.13182/NT70-A28764
Effect of Irradiation on the Elevated Temperature Fracture of Selected Face-Centered Cubic Alloys
M. Kangilaski, S. L. Peterson, J. S. Perrin, R. A. Wullaert
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 550-560
Material | doi.org/10.13182/NT70-A28765
Incoloy 800: Enhanced Resistance to Radiation Damage
D. G. Harman
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 561-571
Material | doi.org/10.13182/NT70-A28766
Reentry Protection for Radioisotope Heat Sources
Thomas S. Bustard, Frank T. Princiotta, Harold N. Barr
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 572-583
Radioisotope | doi.org/10.13182/NT70-A28767
An Analog Computer Controlled Gamma-Ray Spectrometer for Comparative Activation Analysis
P. C. Jurs, T. L. Isenhour
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 584-590
Radioisotopes | doi.org/10.13182/NT70-A28768
Analysis of Gamma-Ray Spectroscopy Data
J. A. Baran, R. S. Reynolds, R. E. Faw, W. R. Kimel
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 591-604
Analysis | doi.org/10.13182/NT70-A28769
Calculation of the Long-Lived Induced Activity in Soil Produced by 200-MeV Protons
T. A. Gabriel
Nuclear Technology | Volume 9 | Number 4 | October 1970 | Pages 605-614
Analysis | doi.org/10.13182/NT70-A28770
Physics Of Operating Boiling Water Reactors
E. D. Fuller
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 622-633
Paper | Reactor | doi.org/10.13182/NT70-A28736
Physics of Operating Pressurized Water Reactors
A. F. McFarlane
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 634-639
Paper | Reactor | doi.org/10.13182/NT70-A28737
Physics Performance of Shippingport
C. A. Flanagan
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 640-650
Paper | Reactor | doi.org/10.13182/NT70-A28738
Experience with Xenon Oscillations
W. E. Graves
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 651-661
Paper | Reactor | doi.org/10.13182/NT70-A28739
Average Thermal Neutron Capture Cross Sections of 198Au, 65Ni, And 66Cu
V. Serment, A. Abu-S Amr A, A. H. Emmons
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 662-666
Paper | Reactor | doi.org/10.13182/NT70-A28740
Spontaneous Deposition of Polonium-210 from Chloride Solution
C. H. H. Chong, M. D. Prisc
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 667-672
Paper | Chemical Processing | doi.org/10.13182/NT70-A28741
Achieving High Exposure in Metallic Uranium Fuel Elements
R. D. Leggett, R. K. Marshall, C. R. Hann, C. H. McGilton
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 673-681
Paper | Fuel | doi.org/10.13182/NT70-A28742
Single- and Two-Phase Pressure Drops on a 16-Rod Bundle
P. Grillo, V. Marinelli
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 682-693
Paper | Fuel | doi.org/10.13182/NT70-A28743
The Application of Reliability Margin Analysis to Fuel Element Performance
C. W. Sayles
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 694-699
Paper | Fuel | doi.org/10.13182/NT70-A28744
Helium Production in EBR-II Irradiated Stainless Steel
N. D. Dudey, S. D. Harkness, H. Farrar, IV
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 700-710
Paper | Fuel | doi.org/10.13182/NT70-A28745
Rate Controlling Factors in the Carbothermic Synthesis of Advanced Fuels
T. B. Lindemer
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 711-715
Fuel | doi.org/10.13182/NT70-A28746
Flowing Sodium Capsules in the GETR
D. L. Brown, G. W. Tunnell
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 716-721
Paper | Material | doi.org/10.13182/NT70-A28747
Interactions Between Radiation Fields from Radioisotope Thermoelectric Generators and Scientific Experiments on Spacecraft
C. G. Miller, V. C. Truscello
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 722-735
Paper | Aerospace | doi.org/10.13182/NT70-A28748
A Neutron Detection System for Operation in Very High Gamma Fields
D. P. Roux, J. T. De Lorenzo
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 736-743
Paper | Instrument | doi.org/10.13182/NT70-A28749
Evaluation Of A Neutron Detection System In A Cobalt-6O Field Of 107 R/H
A. R. Buhl, N. J. Ackermann, Jr., J. T. De Lorenzo
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 744-745
Paper | Instrument | doi.org/10.13182/NT70-A28750
Techniques for Two-Dimensional Gamma-Ray Scanning of Reactor Fuel Element Sections
B. K. Barnes, D. M. Holm, W. M. Sanders, D. D. Clinton, J. E. Swansen
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 746-754
Paper | Technique | doi.org/10.13182/NT70-A28751
Burnup Determination of Nuclear Fuels by High Resolution Gamma Spectrometry, Track Formation in Solid-State Detectors, and Neutron Dose Measurements
P. Popa, M. De Coster, D. Langela
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 755-761
Paper | Technique | doi.org/10.13182/NT70-A28752
Estimation Techniques for Far-field Exposure Contributions
R. S. Reynolds, N. D. Eckhoff
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 762-766
Paper | Technique | doi.org/10.13182/NT70-A28753
Prediction of the Incipient Boiling Conditions Following a Blocked Lmfbr Subassembly Accident
Ralph M. Singer, Robert E. Holtz
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 767-770
Note | Reactor Siting | doi.org/10.13182/NT70-A28754
Ductility Loss in Fast Reactor Irradiated Stainless Steel
A. L. Ward, J. J. Holmes
Nuclear Technology | Volume 9 | Number 5 | November 1970 | Pages 771-772
Note | Material | doi.org/10.13182/NT70-A28755
A Bootstrap Concept of a Safety Test Facility
Charles N. Kelber
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 780-785
Reactor Siting | doi.org/10.13182/NT70-A28709
Civil Defense Implications of a Pressurized Water Reactor in a Thermonuclear Target Area
C. V. Chester, R. O. Chester
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 786-795
Reactor | doi.org/10.13182/NT70-A28710
Evaluation of Isocheck and Invent Fuel Inventory Calculational Models
R. C. Kern, M. V. Bonaca
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 796-806
Fuels Cycle | doi.org/10.13182/NT70-A28711
Uranium-233-Bearing Salt Preparation for the Molten Salt Reactor Experiment
J. M. Chandler, S. E. Bolt
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 807-813
Chemical Processing | doi.org/10.13182/NT70-A28712
An Empirical Formula which Predicts the Critical Parameters of a Planar Array of Uranium-Solution-Filled Cylinders
Harold E. Clark, Grover Tuck
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 814-820
Chemical Processing | doi.org/10.13182/NT70-A28713
The Feasibility of Incorporating Radioactive Wastes in Asphalt or Polyethylene
C. L. Fitzgerald, H. W. Godbee, R. E. Blanco, W. Davis, Jr.
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 821-829
Radioactive Waste | doi.org/10.13182/NT70-A28714
The Synthetic Actinides-From Discovery to Manufacture
Glenn T. Seaborg
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 830-850
Radioisotope | doi.org/10.13182/NT70-A28715
Pulsed-Neutron Activation Analysis System for Short-Lived Radioisotopes
William F. Naughton, William A. Jester
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 851-855
Analysis | doi.org/10.13182/NT70-A28716
Proton Microprobe Analysis of the Surface opf Stranded Wire in the Lunar Module
Gerald M. Padawer
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 856-860
Analysis | doi.org/10.13182/NT70-A28717
Laboratory and Environmental Mineral Analysis using a Californium-252 Neutron Source
R. W. Perkins, L. A. Rancitelli, J. A. Cooper, R. E. Brown
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 861-874
Analysis | doi.org/10.13182/NT70-A28718
A 244Cm-Be Isotopic Neutron Source
D. C. Stewart, E. P. Horwitz, C. H. Youngquist, M. A. Wahlgren
Nuclear Technology | Volume 9 | Number 6 | December 1970 | Pages 875-878
Technique | doi.org/10.13182/NT70-A28719
