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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott 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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Smarter waste strategies: Helping deliver on the promise of advanced nuclear
At COP28, held in Dubai in 2023, a clear consensus emerged: Nuclear energy must be a cornerstone of the global clean energy transition. With electricity demand projected to soar as we decarbonize not just power but also industry, transport, and heat, the case for new nuclear is compelling. More than 20 countries committed to tripling global nuclear capacity by 2050. In the United States alone, the Department of Energy forecasts that the country’s current nuclear capacity could more than triple, adding 200 GW of new nuclear to the existing 95 GW by mid-century.
Yasunori Bessho, Yuichiro Yoshimoto, Osamu Yokomizo, Ryutaro Yamashita, Masumi Ishikawa, Akio Toba
Nuclear Technology | Volume 117 | Number 3 | March 1997 | Pages 281-292
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT97-A35342
Articles are hosted by Taylor and Francis Online.
Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates cal-culational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and a control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bun dle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the onedimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module. In the upper plenum/separator module, the upper plenum is modeled as a single volume in the thermal-equilibrium state and water spiraling in the separator is modeled by an effective length in the momentum equation. In the feedwater/recirculation flow module, the single-phase flow model is solved with the assumption of incompressive flow. Finally, the control system module includes the recirculation flow control minimodule, the pressure control minimodule, and the feedwater control minimodule, as well as the interlock functions, which work during a transient to allow analysis of general transient phenomena. The program was verified to provide satisfactory results within reasonable computational time based on application analysis of stability and scram phenomena in a BWR-5 type plant.
