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Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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.
D. F. Da Cruz, D. Rochman, A. J. Koning
Nuclear Technology | Volume 185 | Number 2 | February 2014 | Pages 174-191
Technical Paper | Fuel Cycle And Management | doi.org/10.13182/NT12-154
Articles are hosted by Taylor and Francis Online.
Uncertainty analysis on reactivity and discharged inventory for a typical pressurized water reactor fuel element as a result of uncertainties in 235,238U, 239,240,241Pu, and fission products nuclear data was performed. A typical Westinghouse three-loop fuel assembly fueled with UO2 fuel with 4.8% enrichment was selected. The Total Monte Carlo method was applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study are from the JEFF3.1 evaluation, with the exception of the nuclear data files for U, Pu, and fission products isotopes (randomized for the generation of the various DRAGON libraries). These are taken from the TALYS evaluated nuclear data library TENDL-2012. Results show that the calculated total uncertainty in keff (as a result of uncertainties in nuclear data of the considered isotopes) is virtually independent of fuel burnup, and amounts to 700 pcm. The uncertainties in the inventory of the discharged fuel are dependent on the element considered and lie in the range 1% to 15% for most fission products, and are <5% for the most important actinides. The total uncertainty on the reactor parameters was also split into different components (different nuclear reaction channels), and the main sources of uncertainties were identified.
