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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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Powering the future: How the DOE is fueling nuclear fuel cycle research and development
As global interest in nuclear energy surges, the United States must remain at the forefront of research and development to ensure national energy security, advance nuclear technologies, and promote international cooperation on safety and nonproliferation. A crucial step in achieving this is analyzing how funding and resources are allocated to better understand how to direct future research and development. The Department of Energy has spearheaded this effort by funding hundreds of research projects across the country through the Nuclear Energy University Program (NEUP). This initiative has empowered dozens of universities to collaborate toward a nuclear-friendly future.
M. K. Moallemi, R. Viskanta
Nuclear Science and Engineering | Volume 98 | Number 3 | March 1988 | Pages 209-225
Technical Paper | doi.org/10.13182/NSE88-A22323
Articles are hosted by Taylor and Francis Online.
A model has been developed to predict the thermal hydraulics in the uncovered part of a pressurized water reactor core. The core is considered to be a heterogeneous porous medium with different permeabilities and effective thermal conductivities in the radial and axial directions. The flow in the core is modeled by the Brinkman-Forchheimer extended Darcy equations. The dependence of the thermophysical properties of the coolant (steam-hydrogen mixture) and the fuel rods with temperature is accounted for. Oxidation of the Zircaloy is also modeled, and transport of the generated hydrogen in the uncovered portion of the reactor core is considered. The effects of the thermal boundary condition at the outlet of the core (i.e., at the upper tie plate) are studied and reported. Partial blockage of the core due to the mechanical failure and/or melting of some of the fuel rods is also modeled, and its effects on the thermal hydraulics of the core are studied and discussed. Numerical simulations are reported for the Three Mile Island Unit 2 reactor conditions. The results show that the flow field in the core is affected by exothermic heat release as well as by a decrease of the coolant density due to the Zircaloy cladding oxidation. In addition, the results show that there is entrapment of the coolant from the upper plenum into the core. The partial blockage of the core was found to have a profound influence on the heatup of the core.