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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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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Latest News
College students help develop waste-measuring device at Hanford
A partnership between Washington River Protection Solutions (WRPS) and Washington State University has resulted in the development of a device to measure radioactive and chemical tank waste at the Hanford Site. WRPS is the contractor at Hanford for the Department of Energy’s Office of Environmental Management.
Akira Yamaguchi, Takashi Takata, Hiroyuki Ohshima, Akikazu Kurihara
Nuclear Technology | Volume 167 | Number 1 | July 2009 | Pages 118-126
Technical Paper | NURETH-12 / Thermal Hydraulics | doi.org/10.13182/NT09-A8856
Articles are hosted by Taylor and Francis Online.
Sodium-water reaction is a design-basis accident of a sodium fast reactor. A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. The typical phenomenon is that the pressurized water blows off and is mixed with the liquid sodium surrounding SG tubes. The design and safety concern is a possibility of the secondary failure of nearby heat transfer tubes that could cause undesirable development of the accident. One needs to evaluate the temperature transients of the heat transfer tubes in the reaction region for safety evaluation. In the present study, a computational method is developed for this purpose. It solves the sodium thermal hydraulics and the heat conduction in the adjacent heat transfer tubes. An experiment performed at the Japan Atomic Energy Agency is analyzed with the method developed in this study. It is found that analyzed temperatures are in good agreement with the experimental data. Based on the experimental and computational results, multiphase multicomponent flow characteristics are depicted. Furthermore, the heat transfer coefficient is evaluated using the instantaneous heat flux and temperature obtained from the numerical simulation.