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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
ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Byoung-Uhn Bae, Jae-Bong Lee, Yu-Sun Park, Jong-Rok Kim, Seok Cho, Kyoung-Ho Kang
Nuclear Technology | Volume 207 | Number 5 | May 2021 | Pages 680-691
Technical Paper | doi.org/10.1080/00295450.2020.1796078
Articles are hosted by Taylor and Francis Online.
To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect test using the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test facility was conducted as the B3.2 test item of the international cooperation project Organisation for Economic Co-operation and Development (OECD)–ATLAS Project Phase 2 (ATLAS-2) (OECD-ATLAS2). The initial and boundary conditions for the test were determined referring to the Advanced Power Reactor 1400 MW(electric) (APR1400) as a prototype with three-level scaling methodology. A single-failure criterion was applied to the operation of the safety injection pump (SIP), and four safety injection tanks (SITs) were available to cool down the reactor coolant system. In the test result, as the break nozzle was located at the DVI line, the clearance of the upper downcomer could make an effective flow path of the steam toward the break and quench the reactor core. Maximum cladding temperature was measured before clearance of the upper downcomer. Coolant inventory in the reactor pressure vessel was maintained due to the safety injection without any further core heatup. So, it was proved that the current design of the safety systems in APR1400 had a sufficient long-term cooling capability with a single SIP during a DVI line break IBLOCA. The ATLAS test data were utilized to evaluate the prediction capability of the thermal-hydraulic system code Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) for a DVI line break IBLOCA scenario. The calculation result with the uncertainty propagation analysis using the PArallel computing Platform IntegRated for Uncertainty and Sensitivity analysis (PAPIRUS) toolkit proved that major phenomena such as uncovery of the core or intermittent injection of the SIT flow could be reasonably predicted by the MARS-KS code.