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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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.
R. Wayne Houston
Nuclear Science and Engineering | Volume 4 | Number 2 | August 1958 | Pages 227-238
doi.org/10.13182/NSE58-A15364
Articles are hosted by Taylor and Francis Online.
For samples exposed in high neutron flux regions of reactors the contribution to the total dosage due to the recoils from elastically scattered fast neutrons may be significant. The calculation of this contribution is considered here. Three methods are presented, each differing in the manner in which the details of the energy distribution of fast neutrons are treated. In the first, the neutron flux per unit energy interval is assumed to be of the asymptotic or 1/E form up to fission energies. In the second and third, a separate computation is made for the uncollided neutrons reaching the sample. The remaining contribution due to once-scattered neutrons is treated as in the first method, but alternate forms for the source spectrum of once-scattered neutrons are considered. Use of the equations requires only a knowledge of the thermal neutron flux in the vicinity of the sample. Assumptions and limitations are discussed. Numerical results are presented for comparison of the effects in light water, heavy water, and graphite moderated reactors in the irradiation of a hydrocarbon (cyclohexane) sample.