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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
Meeting Spotlight
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
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.
Gary E. Giles, Jr.
Nuclear Technology | Volume 117 | Number 3 | March 1997 | Pages 306-315
Technical Paper | Nuclear Fuel Cycle | doi.org/10.13182/NT97-A35345
Articles are hosted by Taylor and Francis Online.
The safety analysis for the Advanced Neutron Source Reactor (ANSR) required the development of a new analysis technique to determine fuel integrity and to assure avoidance of critical heat flux (CHF) conditions. The ANSR is a research reactor design intended to provide the highest continuous neutron beam intensity of any reactor in the world. Reliance on previous safety analysis techniques such as those used in the High Flux Isotope Reactor would result in a design that would not meet the requirements. A more accurate but still conservative analysis technique was developed for the ANSR safety analyses. This technique, the local analysis technique (LAT), relaxed some of the overly conservative assumptions of previous hot-spot studies by using a large number of detailed analyses. The conditions used in these analyses were spread over the possible distributions found in specific designs. This technique was used to analyze several core designs to produce confidence in the fuel plate integrity that could be damaged by excessive fuel temperatures and avoidance of CHF conditions. This approach can be used for other reactor designs and should allow increases in the operating power levels. Alternatively, the LAT could be used to demonstrate increased safety margins for present operating conditions.