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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
Meeting Spotlight
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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
Digital control system installed at China’s Linglong One
Earlier this month, the first digital control system was put in place at Linglong One, a small modular reactor demonstration project being built at the Changjiang nuclear power plant in Hainan Province. This is the world’s first land-based commercial SMR and is controlled by China National Nuclear Power Co. Ltd., a subsidiary of the China National Nuclear Corporation (CNNC).
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.