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Education, Training & Workforce Development
The Education, Training & Workforce Development Division provides communication among the academic, industrial, and governmental communities through the exchange of views and information on matters related to education, training and workforce development in nuclear and radiological science, engineering, and technology. Industry leaders, education and training professionals, and interested students work together through Society-sponsored meetings and publications, to enrich their professional development, to educate the general public, and to advance nuclear and radiological science and engineering.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
Standards Program
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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Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Akio Gofuku, Hidekazu Yoshikawa, Shunsuke Hayashi, Kenji Shimizu, Jiro Wakabayashi
Nuclear Technology | Volume 81 | Number 3 | June 1988 | Pages 313-332
Technical Paper | Fission Reactor | doi.org/10.13182/NT88-A16054
Articles are hosted by Taylor and Francis Online.
A real-time accident tracking calculation technique is investigated for in-depth diagnoses of internal state variables of a pressurized water reactor (PWR) plant in the case of a small-break loss-of-coolant accident (SBLOCA). The technique is composed of two parts: (a) a faster-than-real-time open-system simulator TOKRAC for calculating postaccident thermal-hydraulic behavior in the primary system of a PWR plant, and (b) several real-time state estimators for supplying the unobserved external input to TOKRAC. Observed and unobserved variables are defined as variables that are or are not directly measured by existing plant instrumentation. The external input to the open-system TOKRAC is assumed to be composed of several observed signals and several unobserved state variables: The former are the injection conditions of the emergency core cooling system and the operator actions, while the latter include the total heat transfer rate from the primary to the secondary side in the steam generator (SG) (SG heat transfer rate), the break size and approximate break location in the case of a primary pipe break, or the leak flow rate through an SG tube hole (SG leak rate) in the case of SG tube rupture. The SG heat transfer rate and SG leak rate are first estimated in real time using the observed signals of the SG secondary side. The Kalman filtering technique is applied to a simplified SG secondary model. Due to the estimation technique, the primary side of the PWR plant is separated from the secondary side with respect to heat and mass transfer; then the primary side thermal-hydraulic behavior is rapidly computed using TOKRAC. The unobserved external input, i.e., break size and approximate location of the pipe break in the primary side, is estimated within a short time after the pipe break using the estimated SG heat transfer rate and several observed signals of the primary side. Additional Kalman filters derived by a simplified primary system model are needed to ensure the whole accident tracking calculation. The validity of these estimation techniques is examined through computer experiments. The reference data are taken from the calculation results for several accidents of a Westinghouse-type PWR plant using RELAP4/MOD6 and the experimental data of loss-of-fluid test L9-3. The validity of the whole accident tracking calculation technique is also examined by a computer experiment of a 1.5% cold-leg SBLOCA of a Westinghouse-type PWR plant.