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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.
F. Heidet, R. N. Hill
Nuclear Science and Engineering | Volume 187 | Number 2 | August 2017 | Pages 202-211
Technical Paper | doi.org/10.1080/00295639.2017.1312933
Articles are hosted by Taylor and Francis Online.
Given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235U is replaced with 239Pu or vice versa. These relations are functions of the neutron nonleakage probability, and one only needs to know the number of neutrons emitted per fission and the fission cross-section ratio between the 235U system and the 239Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239Pu as fissile material can be up to 45% larger than when using 235U as fissile material, and the required fissile concentration of 239Pu is up to 48% lower than that of 235U to achieve criticality.