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Division Spotlight
Young Members Group
The Young Members Group works to encourage and enable all young professional members to be actively involved in the efforts and endeavors of the Society at all levels (Professional Divisions, ANS Governance, Local Sections, etc.) as they transition from the role of a student to the role of a professional. It sponsors non-technical workshops and meetings that provide professional development and networking opportunities for young professionals, collaborates with other Divisions and Groups in developing technical and non-technical content for topical and national meetings, encourages its members to participate in the activities of the Groups and Divisions that are closely related to their professional interests as well as in their local sections, introduces young members to the rules and governance structure of the Society, and nominates young professionals for awards and leadership opportunities available to members.
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.
B. S. Pei, Y. B. Chen, Chunkuan Shih, W. S. Lin
Nuclear Technology | Volume 75 | Number 2 | November 1986 | Pages 134-147
Technical Paper | Fission Reactor | doi.org/10.13182/NT86-A33856
Articles are hosted by Taylor and Francis Online.
The performance of five critical heat flux (CHF) correlations with the COBRA IIIC/MIT-1 code was evaluated. These correlations were evaluated against a data group comprised of 2943 axial nonuniform, rod bundle, first-order, and higher rank CHF data points of pressurized water reactor (PWR) core geometries. Among these five CHF correlations, EPRI-1 is the most accurate and has the widest operating ranges. Two kinds of correction factors—cold-wall correction factors and the CHF local quality correction factor—were developed and introduced to EPRI-1 to improve its accuracy in PWR CHF predictions. An in-depth evaluation of the EPRI-1 correlation in the prediction of CHFs of three fuel element abnormalities was also performed. Heat flux spikes and blocked channel conditions have negligible effects on CHFs. For the adverse effects of rod bowing on CHFs, the severity of rod bowing effects depends on the percentage of gap closure between rods, and also on the presence of any thimble tube (cold wall) adjacent to the distorted subchannel. Rod bowing effect parameter correlations under cold-wall conditions were developed. These rod bowing effect parameter correlations were tested; it was proved that they could closely describe the rod bowing effects with no apparent remaining residual trends.
