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Douglas Point Nuclear Generating Station: Not the reactor you may be thinking of
The proposed location of Douglas Point in Maryland, on the banks of the Potomac River, compared to currently operating nuclear plants in Maryland and Virginia.
The Douglas Point Nuclear Generating Station that is the subject of this article is not the CANDU reactor that operated in Ontario from 1966 to 1984. This one was a proposed nuclear power plant in Charles County, Md., that was to provide power to the Washington D.C. area, about 30 miles north of the intended site.
In the early 1970s, the Potomac Electric Power Company (PEPCO) was looking for additional means of generation. At the time, the Washington D.C. metropolitan area was one of the fastest growing regions in the nation.
Site selection was tricky for PEPCO, as the company was contending with a confined load in a growing urban area. A new site as near as possible to the load center that could house at least 2,000 MWe of generating capacity and keep development costs down was needed. Three sites were ultimately reviewed: Douglas Point on the lower Potomac River, a second site toward the mouth of the Potomac River, and a third on the shore of Chesapeake Bay.
A. Bousbia Salah, S. C. Ceuca, R. Puragliesi, R. Mukin, A. Grahn, S. Kliem, J. Vlassenbroeck, H. Austregesilo
Nuclear Technology | Volume 203 | Number 3 | September 2018 | Pages 293-314
Technical Paper | doi.org/10.1080/00295450.2018.1461517
Articles are hosted by Taylor and Francis Online.
Advanced three-dimensional (3-D) computational tools are increasingly being used to simulate complex phenomena occurring during scenarios involving operational transients and accidents in nuclear power plants. Among these scenarios, one can mention the asymmetric coolant mixing under natural-circulation flow regimes. This issue motivated some detailed experimental investigations carried out within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency PKL projects. The aim was not only to assess the mixing phenomenon in the reactor pressure vessel but also to provide experimental data for computer code validations and more specifically thermal-hydraulic system codes with 3-D capabilities. In the current study, the ROCOM/PKL-3 T2.3 experimental test is assessed using, on one hand, thermal-hydraulic system codes with 3-D capabilities and, on the other hand, computational fluid dynamics computational tools. The results emphasize the capabilities and the differences among the considered computational tools as well as their suitability for such purposes.
