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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.
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International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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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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Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
T. J. van Rooyen, G. P. de Beer
Nuclear Science and Engineering | Volume 114 | Number 2 | June 1993 | Pages 87-101
Technical Paper | doi.org/10.13182/NSE93-A24020
Articles are hosted by Taylor and Francis Online.
Prompt fission neutrons (PFNs) constitute the most important component of the source term for nuclear reactor shielding calculations. The determination of the PFN source term for reactor shielding calculations has traditionally been performed using a number of simplifying assumptions. Very simple closed analytical expressions are normally used for the PFN spectrum. The Watt PFN spectrum for 235U, with coefficients determined by Cranberg et al., has become a virtual industry standard in the reactor shielding community. The source term is usually treated as a separable function of spatial location and energy, only the 235U spectrum is considered, and the effect of burnup on the source term is neglected. In reality, the PFN spectra of 235U, 238U, and 239Pu differ markedly, and their fractional contributions to fission are a function of burnup, which, in turn, is a time-dependent function of the spatial position within the reactor core. Recent theoretical developments have led to the advent of sophisticated microscopic models for the calculation of PFN spectra and multiplicities of various fissioning systems. Spectra for 235U, 238U, and 239Pu, calculated with the Madland-Nix model with fragment spin correction, were used in this investigation. An improved reactor source term model that calculates spectrally and spatially burnup-compensated source terms for nuclear reactor shielding calculations is developed and applied to a typical light water reactor (LWR).,Neutron, gamma-ray, and total absorbed dose rate distributions were calculated through four diverse biological shields with a thickness of 250 cm. At end-of-life core conditions, the traditional source term model leads to an underestimate of the transmitted absorbed dose rates by slightly more than a factor of 2. This discrepancy lies within the error margins quoted for LWR shielding calculations. We conclude that despite their age and simplicity, the Watt formula and the simple source term model are of sufficient accuracy for continued service. The more rigorous source term model presented here may be useful for accurate benchmark calculations and for the design of highly efficient shields for high-burnup reactors.
