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The mission of the Decommissioning and Environmental Sciences (DES) Division is to promote the development and use of those skills and technologies associated with the use of nuclear energy and the optimal management and stewardship of the environment, sustainable development, decommissioning, remediation, reutilization, and long-term surveillance and maintenance of nuclear-related installations, and sites. The target audience for this effort is the membership of the Division, the Society, and the public at large.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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ANS announces 2025 Presidential Citations
One of the privileges of being president of the American Nuclear Society is awarding Presidential Citations to individuals who have demonstrated outstanding effort in some manner for the benefit of ANS or the nuclear community at large. Citations are conferred twice each year, at the Annual and Winter Meetings.
ANS President Lisa Marshall has named this season’s recipients, who will receive recognition at the upcoming Annual Conference in Chicago during the Special Session on Tuesday, June 17.
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.