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Division Spotlight
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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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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Latest News
First astatine-labeled compound shipped in the U.S.
The Department of Energy’s National Isotope Development Center (NIDC) on March 31 announced the successful long-distance shipment in the United States of a biologically active compound labeled with the medical radioisotope astatine-211 (At-211). Because previous shipments have included only the “bare” isotope, the NIDC has described the development as “unleashing medical innovation.”
Ketan Ajay, Ravi Kumar, Akhilesh Gupta
Nuclear Science and Engineering | Volume 196 | Number 1 | January 2022 | Pages 75-97
Technical Paper | doi.org/10.1080/00295639.2021.1945393
Articles are hosted by Taylor and Francis Online.
The postulated dual-failure accident, i.e., loss of primary coolant flow along with impairment of the emergency coolant injection system, leads to peak fuel temperatures. It is well known that the temperature of the fuel assemblies is one of the significant factors that affect the outcome of an accident. Therefore, the present work aims to thoroughly investigate the thermal response of a single channel under postulated accident conditions. An experimental system was developed to capture the steady-state heat and temperature distribution in a representative 37-element fuel channel for a decay heat of 6.13 kW. Ohmic heating of the fuel rod simulators (FRSs) mimicked the generation of radioactive decay heat. Numerical simulation was also performed using the Fluent 19.1® code, and the discrete ordinates method was used to solve the radiative transfer equation. Based on the experimental results and the simulation results, it was found that the maximum Zircaloy-4 cladding temperature ≈850°C to 870°C was in the center ring. The temperature was found to vary around the circumference for each of the FRSs. Furthermore, the outer ring FRSs that had the lowest temperature developed the highest circumferential temperature gradient. In the pressure tube, the average circumferential temperature gradient obtained from the experiment and the simulation was 3.76°C/radian and 3.85°C/radian, respectively. Between the calandria tube and the moderator, the heat transfer coefficient was estimated to be around 822.3 W/m2‧K.