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Colin Judge: Testing structural materials in Idaho’s newest hot cell facility
Idaho National Laboratory’s newest facility—the Sample Preparation Laboratory (SPL)—sits across the road from the Hot Fuel Examination Facility (HFEF), which started operating in 1975. SPL will host the first new hot cells at INL’s Materials and Fuels Complex (MFC) in 50 years, giving INL researchers and partners new flexibility to test the structural properties of irradiated materials fresh from the Advanced Test Reactor (ATR) or from a partner’s facility.
Materials meant to withstand extreme conditions in fission or fusion power plants must be tested under similar conditions and pushed past their breaking points so performance and limitations can be understood and improved. Once irradiated, materials samples can be cut down to size in SPL and packaged for testing in other facilities at INL or other national laboratories, commercial labs, or universities. But they can also be subjected to extreme thermal or corrosive conditions and mechanical testing right in SPL, explains Colin Judge, who, as INL’s division director for nuclear materials performance, oversees SPL and other facilities at the MFC.
SPL won’t go “hot” until January 2026, but Judge spoke with NN staff writer Susan Gallier about its capabilities as his team was moving instruments into the new facility.
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Automatic Rapid Process for the Generation of Problem-Dependent SAS2H/ORIGEN-S Cross-Section Libraries
Luiz C. Leal, Otto W. Hermann, Stephen M. Bowman, Cecil V. Parks
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 1-23
Technical Paper | Fission Reactors | doi.org/10.13182/NT99-A2980
Drag Coefficients for Reactor Loose Parts
Liang Shi, J. Michael Doster, Charles W. Mayo
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 24-37
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2981
Modeling the Statics of a Natural-Circulation Boiling Water Reactor Loop
Haico V. Kok, Tim H. J. J. van der Hagen
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 38-48
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2982
Qualification of a Three-Dimensional Core Dynamics Analysis Program Coupled with a Detailed Mesh Division for Commercial Boiling Water Reactors-I
Yasunori Bessho, Takafumi Anegawa, Osamu Yokomizo, Yuichiro Yoshimoto, Masao Chaki, Motoo Aoyama, Takanori Fukahori
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 49-65
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2983
Natural Convection Heat Transfer with Crust Formation in the Molten Metal Pool
Rae-Joon Park, Sang-Baik Kim, Hee-Dong Kim, Sang-Min Choi
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 66-80
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2984
Separation of Radioactive Strontium from Alkaline Nuclear Waste Solutions with the Highly Effective Ion Exchanger SrTreat
Jukka Lehto, Leena Brodkin, Risto Harjula, Esko Tusa
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 81-87
Technical Paper | Radioactive Waste Management and Disposal | doi.org/10.13182/NT99-A2985
Dose Rate Measurements and Calculation of TN-12/2 Packages
Hiroaki Taniuchi, Fumio Matsuda
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 88-101
Technical Paper | Radioactive Waste Management and Disposal | doi.org/10.13182/NT99-A2986
Empirical Steam Generator Water-Level Modeling Using Neural Networks
Chaung Lin, Tsung-Ming Lin
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 102-112
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT99-A2987
Theoretical Calculations of Proton Reaction Cross Sections in a Lead Target with Energy Ep 300 MeV
Jinfeng Zhou, Qingbiao Shen, Xiuquan Sun
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 113-122
Technical Paper | Radiation Measurements and Instrumentation | doi.org/10.13182/NT99-A2988
Uncertainty Analysis in DUPIC Fuel-Cycle Cost Using a Probabilistic Simulation Method
Won Il Ko, Jong Won Choi, Jae Sol Lee, Hyun Soo Park, Kun Jai Lee
Nuclear Technology | Volume 127 | Number 1 | July 1999 | Pages 123-140
Technical Note | Fuel Cycle and Management | doi.org/10.13182/NT99-A2989
A MELCOR-Based Model to Calculate Vapor Natural Circulation in Pressurized Water Reactor High-Pressure Severe Accidents
Francisco Martín-Fuertes Hernández-Sonseca
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 141-150
Technical Paper | Reactor Safety | doi.org/10.13182/NT99-A2990
Modeling of the Rim Effect on Thermal Behavior of High-Burnup UO2 Fuel
Byung-Ho Lee, Yang-Hyun Koo, Dong-Seong Sohn
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 151-159
Technical Paper | Fuel Cycle and Management | doi.org/10.13182/NT99-A2991
A Condensation Experiment in the Presence of Noncondensables in a Vertical Tube of a Passive Containment Cooling System and Its Assessment with RELAP5/MOD3.2
Hyun Sik Park, Hee Cheon No
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 160-169
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2992
Core Damage Severity Evaluation for Pressurized Water Reactors by Use of Fuzzy Logic Algorithms
Anastasios Mironidis, Leon Lidofsky, George Grochowski, Lefteris Tsoukalas
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 170-185
Technical Paper | Reactor Operations and Control | doi.org/10.13182/NT99-A2993
Conceivable Limits for Thermal Reactor Spent-Fuel Recycling
David R. Kingdon, Vladimir A. Khotylev, Archie A. Harms, J. Eduard Hoogenboom
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 186-198
Technical Paper | Reprocessing | doi.org/10.13182/NT99-A2994
Matrix Diffusion of Simple Cations, Anions, and Neutral Species in Fractured Crystalline Rocks
Haruo Sato
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 199-211
Technical Paper | Radioactive Waste Management and Disposal | doi.org/10.13182/NT127-199
Evaluation of the Effect of Temperature on Potassium Clinoptilolite Ion Exchange
David A. White, Andreas Nattkemper, Radu Ratautiu
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 212-217
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT99-A2996
Design of the European Spallation Source Liquid-Metal Target Using Computational Fluid Dynamics
Trevor V. Dury, Brian L. Smith, Günter S. Bauer
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 218-232
Technical Paper | Accelerators | doi.org/10.13182/NT99-A2997
An Electrocatalyzed Process for Leaching of Plutonium from Active Gauntlets and Subsequent Separation by Solvent Extraction
K. Venugopal Chetty, P. M. Mapara, A. G. Godbole, Rajendra Swarup
Nuclear Technology | Volume 127 | Number 2 | August 1999 | Pages 233-237
Technical Note | Reprocessing | doi.org/10.13182/NT99-A2998
Feasibility Studies of a Soluble Boron-Free 900-MW(electric) PWR, Safety Systems: Consequences of the Partial or Total Elimination of Soluble Boron on Plant Safety and Plant Systems Architecture
Gian Luigi Fiorini, Guy Marie Gautier, Yves Bergamaschi
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 239-258
Technical Paper | Fission Reactors | doi.org/10.13182/NT99-A2999
Feasibility Studies of a Soluble Boron-Free 900-MW(electric) PWR, Core Physics - I: Motivations, Assembly Design, and Core Control
Paul Thomet
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 259-266
Technical Paper | Fission Reactors | doi.org/10.13182/NT127-259
Feasibility Studies of a Soluble Boron-Free 900-MW(electric) PWR, Core Physics - II: Control Rod Follow, Load Follow, and Reactivity-Initiated Accident Linked to RCCAs
Jean-Paul Deffain, Philippe Alexandre, Paul Thomet
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 267-286
Technical Paper | Fission Reactors | doi.org/10.13182/NT99-A3001
Hybrid Analysis of the Simplified Boiling Water Reactor Using RAMONA-4B and CASMO-3 Computer Codes
Gabriel F. Cuevas Vivas, Yassin A. Hassan
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 287-300
Technical Paper | Reactor Safety | doi.org/10.13182/NT99-A3002
Benchmark Calculations of the ESADA Single-Region Mixed-Oxide Critical Experiments
Hatice Akkurt, Naeem M. Abdurrahman
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 301-314
Technical Paper | Reactor Safety | doi.org/10.13182/NT99-A3003
Benchmark Calculations of the Saxton Plutonium Program Critical Experiments
Naeem M. Abdurrahman, Georgeta Radulescu, Igor Carron
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 315-331
Technical Paper | Reactor Safety | doi.org/10.13182/NT99-A3004
Behavior of a UF6 Container During a Fire - I: Analysis and Phenomenological Interpretation of Tenerife Experimental Results
Eric Pinton, Bernard Duret, Georges Berthoud
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 332-351
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A3005
Behavior of a UF6 Container in a Fire - II: Modeling and Validation
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 352-374
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A3006
Uranium Dioxide Powder Preparation, Pressing, and Sintering for Optimum Yield
Palanki Balakrishna, Chandru K. Asnani, R. M. Kartha, K. Ramachandran, K. Sarat Babu, Vaidyanathan Ravichandran, Bhallamudi Narasimha Murty, Chaitanyamoy Ganguly
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 375-381
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT99-A3007
Pressurized Water Reactor Coolant Pump Locked Rotor Analysis Methodology Using RETRAN-02 and FREY
Lainsu Kao, Ping-Hue Huang
Nuclear Technology | Volume 127 | Number 3 | September 1999 | Pages 382-388
Technical Note | Thermal Hydraulics | doi.org/10.13182/NT99-A3008
