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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.”
Sule Ergun, Jason G. Williams, Lawrence E. Hochreiter, Hergen Wiersema, Marcel Slootman, Marek Stempniewicz
Nuclear Technology | Volume 156 | Number 1 | October 2006 | Pages 69-74
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT06-A3774
Articles are hosted by Taylor and Francis Online.
Critical heat flux (CHF) at a natural boiling condition is an important phenomenon for a research reactor having a small-hydraulic-diameter geometry under a large-break loss-of-coolant accident condition. Accurately predicting the CHF under this condition is very important; therefore, the CHF models used in the best-estimate codes must be validated using appropriate experimental data for a given geometry. The present work focuses on validating the CHF calculations and models within the COolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code by simulating two sets of experiments, which were performed in tubes and annuli with different length-to-diameter ratios. In this work, the cocurrent upflow and downflow correlations developed by Mishima and Nishihara and Holowach et al. and Zuber correlations for the CHF used in COBRA-TF are validated against the experimental data obtained by Monde and Yamaji and Islam et al. Conclusions on the predictive capability of COBRA-TF for the CHF calculations for small-hydraulic-diameter geometry under natural boiling conditions are provided with the description of the correlations and models used.
