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Fusion Energy
This division promotes the development and timely introduction of fusion energy as a sustainable energy source with favorable economic, environmental, and safety attributes. The division cooperates with other organizations on common issues of multidisciplinary fusion science and technology, conducts professional meetings, and disseminates technical information in support of these goals. Members focus on the assessment and resolution of critical developmental issues for practical fusion energy applications.
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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.
Borut Mavko, Andrej Prošek, Francesco D’auria
Nuclear Technology | Volume 120 | Number 1 | October 1997 | Pages 1-18
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT97-A35427
Articles are hosted by Taylor and Francis Online.
Quantitative evaluation of thermal-hydraulic code uncertainties is a necessary step in the code assessment process, especially if best-estimate codes are utilized for licensing purposes. With the goal of quantifying code accuracy, researchers in the past developed a methodology based on the fast Fourier transform (FFT) that consisted of qualitative and quantitative code assessment. Here, the FFT-based method is applied to International Atomic Energy Agency (IAEA)-Standard Problem Exercise (SPE)-4 test results with pre- and posttest code calculations of the IAEA-SPE-4 experiment. Four system codes (ATHLET, CATHARE, MELCOR, and RELAP5) are used for calculations of the experiment, performed at the PMK-2 facility, which simulated a cold-leg break in a WER-440 plant. The results show that the posttest calculations had better accuracy than did the pretest calculations. None of the best three pre- and posttest calculations were able to predict core dryout, which was the most important phenomenon observed during the test. The results obtained can give an objective indication of the capability of the aforementioned codes in predicting relevant variables characterizing the transient (too few experimental parameters may limit full application of the FFT-based methodology).