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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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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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NN Asks: What did you learn from ANS’s Nuclear 101?
Mike Harkin
When ANS first announced its new Nuclear 101 certificate course, I was excited. This felt like a course tailor-made for me, a transplant into the commercial nuclear world. I enrolled for the inaugural session held in November 2024, knowing it was going to be hard (this is nuclear power, of course)—but I had been working on ramping up my knowledge base for the past year, through both my employer and at a local college.
The course was a fast-and-furious roller-coaster ride through all the key components of the nuclear power industry, in one highly challenging week. In fact, the challenges the students experienced caught even the instructors by surprise. Thankfully, the shared intellectual stretch we students all felt helped us band together to push through to the end.
We were all impressed with the quality of the instructors, who are some of the top experts in the field. We appreciated not only their knowledge base but their support whenever someone struggled to understand a concept.
Siyang Huang, Qiqi Yan, Wenxi Tian, G. H. Su, Suizheng Qiu (Xi’an Jiaotong Univ)
Proceedings | Advances in Thermal Hydraulics 2018 | Orlando, FL, November 11-15, 2018 | Pages 44-56
In the nuclear power system, the critical heat flux (CHF) plays a crucial role in the reactor safety analysis. When CHF occurs, it will cause a sudden increase in the surface temperature, which would lead to the failure of fuel claddings and damage of the core. Considering the cross flow between neighboring channels, spacer grids and mixing vanes in the fuel assembly, the local flow conditions and the geometry of the flow channels make the prediction of CHF more complicated. In this paper, the departure from nucleate boiling (DNB) type CHF in rod bundle is investigated based on the coupled analysis of the subchannel method and a CHF mechanism model, i.e. the liquid sublayer dryout model. The liquid sublayer dryout model assumes that there is a thin liquid sublayer underneath a vapor blanket formed by the coalescence of small bubbles near the heated wall. The dryout of this sublayer will be regarded as the CHF occurrence. In present research, the homogeneous flow model is adopted in the subchannel analysis code to predict the local flow conditions for the rod bundle flow subchannels, which will be used as the input parameters for the liquid sublayer dryout model. In order to verify the method above, the predicted results are compared with the CHF Look-Up Table 2006 (LUT-2006) and the predicted results are in good agreement with the data in LUT-2006. In addition, the effects of rod bundle inlet subcooling, mass flux, heated length and motion conditions on CHF are analyzed.
