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Fusion Energy
This division promotes the development and timely introduction of fusion energy as a sustainable energy source with favorable economic, environmental, and safety attributes. The division cooperates with other organizations on common issues of multidisciplinary fusion science and technology, conducts professional meetings, and disseminates technical information in support of these goals. Members focus on the assessment and resolution of critical developmental issues for practical fusion energy applications.
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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.
Theron D. Marshall, Robert W. Hockenbury, John A. Honey, Lee C. CadWallader
Nuclear Technology | Volume 114 | Number 1 | April 1996 | Pages 84-96
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT96-A35225
Articles are hosted by Taylor and Francis Online.
Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. U.S.-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here were taken from a U.S. Nuclear Regulatory Commission summary report on severe accident risk: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, arid small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability datafrom U. S. PWRs and (b) assuming common component reliability data. U.S. PWR component reliability data are used because of the unavailability of such data for the V213 at the time of the analyses. While the use of U.S. PWR data in this risk assessment of the V213 does strongly infer V213 comparability to U.S. plants, the risk assessment using common component reliability does not have such a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor’s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to U.S. PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures. Additionally, the analyses make no implications as to the effects of the selected initiating events on the health and safety of the public.