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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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A Computerized Diagnostic System for Nuclear Plant Control Rooms Based on Statistical Quality Control
Carolyn D. Heising, William S. Grenzebach
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 7-15
Technical Paper | Fission Reactor | doi.org/10.13182/NT90-A34381
Iodine Partitioning in Pressurized Water Reactor Steam Generator Accidents
Edward C. Beahm, Steven R. Daish, William E. Shockley, Joram Hopenfeld
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 16-22
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34382
Measurements of Radioiodine Species in Samples of Pressurized Water Reactor Coolant
Paul G. Voillequé
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 23-33
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34383
Automation of Nuclear Power Plants
Abdo A. Husseiny, Zeinab A. Sabri, S. Keith Adams, Rodrigo J. Rodriguez
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 34-48
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34384
Assessment of the Safety Function for the Anticipated Transient without Trip Mitigation System Actuation Circuitry at Maanshan Nuclear Power Station
Bau-Shei Pei, Ge-Ping Yu, Guei-Ching Lin, Yin-Pang Ma
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 49-62
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34385
Assessment of Uncertainties in Early Off-Site Consequences from Nuclear Reactor Accidents
Imtiaz K. Madni, Erik G. Cazzoli, Mohsen Khatib-Rahbar
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 63-77
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34386
Measuring and Predicting Gamma Radiation from Radioactive Glass-Filled Canisters
Richard D. Peters, Urban P. Jenquin, Langdon K. Holton, Jr.
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 78-86
Technical Paper | Radioactive Waste Management | doi.org/10.13182/NT90-A34387
Studies on the Applicability of a Flow Coupler to a Liquid-Metal Fast Breeder Reactor Plant
Sadao Hattori, Tadasu Takuma, Koshichi Nemoto, Masafumi Terada, Tamotsu Sano
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 87-97
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34388
UTSG-2—A Theoretical Model Describing the Transient Behavior of a Pressurized Water Reactor Natural-Circulation U-Tube Steam Generator
Alois Höld
Nuclear Technology | Volume 90 | Number 1 | April 1990 | Pages 98-118
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34389
A Distribution Parameter Derived for Rectangular Channels and Simulated Subchannel Geometry
Hasna J. Khan
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 125-141
Technical Paper | Fission Reactor | doi.org/10.13182/NT90-A34409
A Three-Dimensional Space-Time Model and Its Use in Pressurized Water Reactor Rod Ejection Analyses
H. P. Chou, J. R. Lu, M. B. Chang
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 142-154
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34410
Event Sequence of a Severe Accident in a Single-Unit Candu Reactor
Jerry E. Dick, Vijay I. Nath, Erl Kohn, Thomas K. Min, Soedi Prawirosoehardjo
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 155-167
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34411
Frequency and Consequences Associated with a Steam Generator Tube Rupture Event
James P. Adams, Martin B. Sattison
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 168-185
Technical Paper | Nuclear Safety | doi.org/10.13182/NT90-A34412
Nuclear Fuel Reprocessing of (U,Pu)02 Fuel
Akihiko Inoue
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 186-190
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT90-A34413
A Nondestructive Method for Light Water Reactor Fuel Assembly Identification
Hermann Würz, Werner Eyrich, Hans-Joachim Becker
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 191-204
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT90-A34414
Brine Migration in a Salt Repository
Yongsoo Hwang, P. L. Chambré, T. H. Pigford, W. W.-L. Lee
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 205-214
Technical Paper | Radioacitive Waste Management | doi.org/10.13182/NT90-A34415
Derivation of an Equation for Radionuclide Transport in Porous Media
Fu-Long Chen, Shih-Hai Li
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 215-225
Technical Paper | Radioacitive Waste Management | doi.org/10.13182/NT90-A34416
Reaction Behavior of B4C Absorber Material with Stainless Steel and Zircaloy in Severe Light Water Reactor Accidents
Peter Hofmann, Mario Enrique Markiewicz, José Luis Spino
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 226-244
Technical Paper | Matetial | doi.org/10.13182/NT90-A34417
A Generalized Methodology for Obtaining Quantitative Charpy Data from Test Specimens of Nonstandard Dimensions
Michael P. Manahan, Sr., Christopher Charles
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 245-259
Technical Paper | Material | doi.org/10.13182/NT90-A34418
Education and Training at the Pennsylvania State University Breazeale Reactor Over the Past 20 Years
Samuel H. Levine, Marcus H. Voth
Nuclear Technology | Volume 90 | Number 2 | May 1990 | Pages 260-265
Technical Paper | Education | doi.org/10.13182/NT90-A34419
Modeling and Loss-of-Coolant Accident Analysis of a Nuclear Power Plant Using RELAP5/MOD2
Parvez Salim, Yassin A. Hassan
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 275-285
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34393
Pump Cavitation in Savannah River Reactors During Loss-of-Coolant Accidents
Cliff B. Davis
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 286-293
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34394
Transient Analysis in the Ascó Nuclear Power Plant Using RELAP5/MOD2
Francesc Reventós, José Sánchez-Baptista, Alberto Pérez Navas, Pablo Moreno
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 294-307
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34395
Application of RELAP5/MOD2 to Loviisa Nuclear Power Plant Overcooling Transients
Heikki Kantee, Harri Tuomisto, Vesa Yrjölä
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 308-315
Technical Paper | RELAP/MOD2 / Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34396
The RELAP5-FORCE MOD2 Code, A Hydrodynamic Forcing Function Calculation Version of RELAP5
Juan M. Cajigas
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 316-325
Technical Paper | RELAP/MOD2 / Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34397
A Comparison Study of the Westinghouse Model E Steam Generator Using RELAP5/MOD2 and RETRAN-02 Computer Codes
Thomas K. Blanchat, Yassin A. Hassan
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 326-339
Technical Paper | RELAP/MOD2 / Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34398
Assessment of the RELAP5/MOD2 Code on the Basis of Experiments Performed in the LOBI Facility
Francesco D'auria, Giorgio Maria Galassi
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 340-355
Technical Paper | RELAP/MOD2 / Heat Transfer and Fluid Flow | doi.org/10.13182/NT90-A34399
FISA-2/WS, A Compact Real-Time Simulator for Two-Loop Pressurized Water Reactor Plants
Jae Jun Jeong, Deog Yeon Oh, Hee Cheon No, Soon Heung Chang, Sung Jae Cho, Hwang Yong Jun, Yong Kwan Lee
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 356-370
Technical Paper | RELAP/MOD2 / Fission Reactor | doi.org/10.13182/NT90-A34400
An analysis of the Physical Causes of the Chernobyl Accident
José M. MartíNez-Val, José M. AragonéS, Emilio míNguez, José M. Perlado, Guillermo Velarde
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 371-388
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34401
Optimization of Eddy-Current Probes for Detection of Garter Springs in Pressurized Heavy Water Reactors
Bhagi Purna Chandra Rao, Mandayam Tondanur Shyamsunder, Dipak Kumar Bhattacharya, Baldev Raj
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 389-393
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34402
Experimental Verification of Hygroscopic Aerosol Growth in Reactor Accident Conditions
Jorma Jokiniemi, Kimmo Koistinen, Taisto Raunemaa
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 394-407
Technical Paper | RELAP/MOD2 / Nuclear Safety | doi.org/10.13182/NT90-A34403
An Electrochemical Hydrogen Meter for Measurement of Dissolved Hydrogen in Liquid Sodium
Thiagarajan Gnanasekaran, Kandhalu Hari Mahendran, Raghavachary Sridharan, Vedaraman Ganesan, Govindaswami Periaswami, Cherian K. Mathews
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 408-416
Technical Paper | RELAP/MOD2 / Material | doi.org/10.13182/NT90-A34404
A Comparison of Prediction and Tapucu Experimental Data for Determination of the Axial Velocity in the Gap Region
A. Cihat Baytaṣ
Nuclear Technology | Volume 90 | Number 3 | June 1990 | Pages 417-425
Technical Paper | RELAP/MOD2 / Heat Transfer and fluid Flow | doi.org/10.13182/NT90-A34405
