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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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Remembering ANS member Gil Brown
Brown
The nuclear community is mourning the loss of Gilbert Brown, who passed away on July 11 at the age of 77 following a battle with cancer.
Brown, an American Nuclear Society Fellow and an ANS member for nearly 50 years, joined the faculty at Lowell Technological Institute—now the University of Massachusetts–Lowell—in 1973 and remained there for the rest of his career. He eventually became director of the UMass Lowell nuclear engineering program. After his retirement, he remained an emeritus professor at the university.
Sukesh Aghara, chair of the Nuclear Engineering Department Heads Organization, noted in an email to NEDHO members and others that “Gil was a relentless advocate for nuclear energy and a deeply respected member of our professional community. He was also a kind and generous friend—and one of the reasons I ended up at UMass Lowell. He served the university with great dedication. . . . Within NEDHO, Gil was a steady presence and served for many years as our treasurer. His contributions to nuclear engineering education and to this community will be dearly missed.”
Richard R. Trewin
Nuclear Technology | Volume 208 | Number 5 | May 2022 | Pages 860-870
Technical Paper | doi.org/10.1080/00295450.2021.1964312
Articles are hosted by Taylor and Francis Online.
The ability to extend the operating life of a pressurized water reactor depends in part on the ability of the reactor pressure vessel to withstand thermal shock concurrent with significant pressure. If during the course of a small-break loss-of-coolant accident (SBLOCA), the primary-side pressure is reduced sufficiently, cold make-up water is supplied to the cold leg by the emergency core cooling system. If incomplete mixing occurs between the cold injected water and the hot water in the primary circuit, a stream of cool water flows along the bottom of the cold leg into the downcomer. There, the cool water forms a downward-flowing buoyant plume surrounded by the hot water in the downcomer. The time-dependent spatial distributions of the temperatures and heat transfer coefficients on the inside surface of the reactor pressure vessel are important in determining compliance with regulatory requirements. The simulation of the mixing in the cold leg and downcomer is typically performed with flow-mixing computer codes, most of which use either computational fluid dynamics techniques or mechanistic models. The computer code used for this work, called KWU-MIX, makes use of mechanistic models. In previous works, the uncertainties in parameters associated with the most important phenomena that contribute to the temperature distributions were quantified by comparing experimentally derived values of the parameters with values from the mechanistic models. In this work, those uncertainties are propagated through the flow-mixing code in order to quantify the uncertainty in the calculated temperature distributions. An example of the propagation of uncertainties is given for conditions typical of a SBLOCA. Random values from each of the uncertainty distributions for the parameters of all of the most important phenomena were selected for each of 100 simulations of the typical accident conditions. The results of the 100 simulations were analyzed statistically in order to quantify the best-estimate temperature distribution and its uncertainty. The resulting best-estimate temperature distribution and its uncertainty were compared with experimental data obtained in the Upper Plenum Test Facility at the same typical accident conditions. The results of the comparison show that the uncertainty in the calculated temperature distribution bounds the experimental values.