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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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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Kazuyuki Takase, Yasuo Ose, Hajime Akimoto
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 1050-1055
Safety and Environment | doi.org/10.13182/FST01-A11963382
Articles are hosted by Taylor and Francis Online.
Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.
The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.