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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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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Crash Course: The DOE’s Package Performance Demonstration
Inspired by a history of similar testing endeavors and recommended by the National Academy of Sciences and the Blue Ribbon Commission on America’s Nuclear Future, the Department of Energy is planning to conduct physical demonstrations on rail-sized spent nuclear fuel transportation casks. As part of the project, called the Spent Nuclear Fuel Package Performance Demonstration (PPD), the DOE is considering a number of demonstrations based on regulatory tests and realistic transportation scenarios, including collisions, drops, exposure to fire, and immersion in water.
Kwang Soon Ha, Fan-Bill Cheung, Jinho Song, Rae Joon Park, Sang Baik Kim
Nuclear Technology | Volume 181 | Number 1 | January 2013 | Pages 196-207
Technical Paper | Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Thermal Hydraulics | doi.org/10.13182/NT13-A15767
Articles are hosted by Taylor and Francis Online.
Boiling-induced natural-circulation flow in various engineered cooling channels is modeled and solved by considering the conservation of mass, momentum, and energy in the two-phase mixture, along with the two-phase friction drop and void fraction. The model is applied to estimate the induced mass flow rates through a uniform annular gap and a nonuniform annular gap between the reactor vessel and insulation under the in-vessel corium retention-external reactor vessel cooling conditions, and in the engineered corium cooling system of an ex-vessel core catcher during a severe accident. Dependence of the induced flow rate on various system parameters including the channel gap size, inlet diameter, inlet subcooling, and wall heat flux has been identified numerically. Results of the present study provide useful information for enhancing the design of engineered cooling channels to assure long-term cooling and retention of corium under severe accident conditions.