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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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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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Latest News
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.
Te-Chuan Wang, Shih-Jen Wang, Jyh-Tong Teng
Nuclear Technology | Volume 152 | Number 3 | December 2005 | Pages 253-265
Technical Paper | Fission Reactors | doi.org/10.13182/NT05-A3674
Articles are hosted by Taylor and Francis Online.
The Chinshan nuclear power plant (NPP) is a Mark-I boiling water reactor (BWR) NPP located in northern Taiwan. The Chinshan NPP severe accident management guidelines (SAMGs) were developed based on the BWR Owners Group Emergency Procedure Guidelines/Severe Accident Guidelines and were developed at the end of 2003. The MAAP4 code has been used as a tool to validate the SAMG strategies. The development process and characteristics of the Chinshan SAMGs are described. The T5UtXC sequence, the highest core damage frequency in the probabilistic risk assessment insight of the Chinshan NPP, is cited as a reference case for SAMG validation. Not all safety injection systems are operated in the T5UtXC sequence. The severe accident progression is simulated, and the entry condition of the SAMGs is described. Then, the T5UtXC sequence is simulated with reactor pressure vessel (RPV) depressurization. Mitigation actions based on the Chinshan NPP SAMGs are then applied to demonstrate the effectiveness of the SAMGs. Sensitivity studies on RPV depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. Based on MAAP4 calculation and the default values of the parameters calculating the severe accident phenomena, the result shows that RPV depressurization before the reactor water level reaches one-fourth of the core water level can prevent the core from damage in the T5UtXC sequence. The flow rate of two control rod drive pumps is enough to maintain the reactor water level above the top of active fuel and cool down the core in the T5UtXC sequence without operator action.
