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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.”
Kostadin Ivanov, Enrico Sartori, E. Royer, S. Langenbuch, K. Velkov
Nuclear Technology | Volume 157 | Number 2 | February 2007 | Pages 177-195
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT07-A3811
Articles are hosted by Taylor and Francis Online.
Incorporating full three-dimensional models of the reactor core into system transient codes allows for a "best-estimate" calculation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations on the development of coupled thermal-hydraulic and neutronics codes. Appropriate benchmarks have been developed in international cooperations led by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that permit testing of the neutronics-thermal-hydraulics coupling and verification of the capability of the coupled codes to analyze complex transients with coupled core-plant interactions. Three such benchmarks are presented in this paper - the OECD/U.S. Nuclear Regulatory Commission (NRC) pressurized water reactor main steam line break benchmark, the OECD/NRC boiling water reactor turbine trip benchmark, and the OECD/U.S. Department of Energy/Commissariat à l'Energie Atomique V1000 coolant transient benchmark. To meet the objectives of the validation of best-estimate coupled codes, a systematic approach has been introduced to evaluate the analyzed transients employing a multilevel methodology. Since these benchmarks are based on both code-to-code and code-to-data comparisons, further guidance for presenting and evaluating results has been developed. During the course of the benchmark activities, a professional community has been established, which allowed our carrying out in-depth discussions of different aspects considered in the validation process of the coupled codes. This positive output has certainly advanced the state of the art in the area of coupling research.