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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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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
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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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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.
Nuclear Technology | Volume 93 | Number 1 | January 1991 | Pages 53-64
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT91-A34518
Articles are hosted by Taylor and Francis Online.
A postulated loss-of-residual heat removal (RHR) event for a reactor coolant system (RCS) in a midloop condition is analyzed for the Palo Verde Nuclear Generating Station (PVNGS) using the RETRAN-2/MOD4 computer code. The PVNGS is a Combustion Engineering, two-loop, 3800-MW(thermal) pressurized water reactor (PWR). This analysis was prompted by the U.S. Nuclear Regulatory Commission generic letter 88-17, which requires PWR licensees to perform analyses of loss-of-RHR events for their facilities. Such analyses yield a detailed understanding of loss-of-RHR events and provide a basis for emergency procedures and recovery actions. Simulations of events following a loss of RHR are used to determine the time for RCS coolant to reach boiling, evaluate the potential for a rapid core uncovery due to pressurization of the reactor ves sel head with a corresponding liquid ejection from a cold-leg breach, and evaluate options for maintaining the RCS inventory above the core. Key features of the PVNGS midloop condition RETRAN model include isolating both steam generators from the RCS by nozzle dams, setting the initial RCS inventory at midloop, venting the RCS to containment by removing the pressurizer safety relief valves, defining a maintenance breach on one cold leg, and providing options to evaluate makeup water injection to a hot or cold leg by either pump or gravity feed. The RETRAN model provides a satisfactory method for dynamically evaluating loss-of-RHR events from a reduced RCS inventory condition, and for evaluating alternative recovery actions for PVNGS.