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Conference Spotlight
2025 ANS Winter Conference & Expo
November 9–12, 2025
Washington, DC|Washington Hilton
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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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.
Junghyun Bae, Robert S. Bean
Nuclear Science and Engineering | Volume 196 | Number 10 | October 2022 | Pages 1224-1235
Technical Paper | doi.org/10.1080/00295639.2022.2055700
Articles are hosted by Taylor and Francis Online.
In pool-type research reactors, the fuel core is placed in a large open pool of water, and it is consistently cooled by natural circulation. To meet the increasing demands of reactor-based research, i.e., neutron irradiation and isotope production, many institutes have been considering upgrading the designed power levels of their research reactors to maximize their utility. However, increasing operating power levels without replacing the major components of the reactor system is challenging because two important analyses must be extensively performed: (1) neutron transport analysis for nuclear fission and decay heat generation and (2) thermohydraulic analysis for heat removal in the core. In this paper, we investigate thermohydraulic limits on the maximum power of the Purdue University research reactor (PUR-1) using computational fluid dynamics (CFD) simulations which are coupled with the results from Monte Carlo neutron transport simulations. We design a PUR-1 fuel assembly, which is designated as the hottest one for CFD simulations, that includes a narrow, rectangular, and upward coolant channel. Here we demonstrate that the thermohydraulic limit for PUR-1 core power is 350 kW without changing the coolant system. Given a conservative safety margin, however, the estimated maximum power level is decreased to 170 kW. In the end, the results of two additional cooling systems—guide pipe and lowered coolant temperature—are presented to demonstrate the potential of advanced cooling capacity. They would enable reactors to operate at higher core power levels.