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Division Spotlight
Robotics & Remote Systems
The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
Meeting Spotlight
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
Standards Program
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.
D. Shome, M. A. R. Sarkar (BUET)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 893-899
The objective of this paper is to present and analyze the results of simulated tube rupture accident in VVER-1000 Nuclear Reactor in PCTRAN. In simulating the accident, 100% of one full tube rupture has been considered. The simulation result shows that the core pressure experience a rapid decrease from initial value of 155 bar (15.5 MPa) and stabilize around 80 bar (8 MPa) after the accident. This leads to stopping coolant leakage from primary circuit to secondary circuit due to absence of pressure differential between primary and secondary loop. After the initiation of tube rupture, the leak from affected Steam Generator ‘A’ is about 3000 t/h (833.33 kg/s) which is reduced to approximately 500 t/h(138.89kg/s) within 200s of the accident. The result also shows that the reactor power (both ‘Thermal’ and ‘Nuclear Flux’) collapses drastically following reactor trip. Both High Pressure Injection (HPI) pump is activated following “Reactor Scram” to prevent core damage. The average temperature of coolant at the reactor inlet decreases from 580K to 560K to facilitate cooling down of the primary coolant. The data obtained from the simulation are satisfactorily consistent with PSAR (Preliminary Safety Assessment Report) data regarding SGTR accident. These findings are expected to provide useful information in understanding and evaluating plants capability to mitigate the consequence of SGTR accident.