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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.”
Dong Won Lee, Bong Geun Hong, Yonghee Kim, Wang Ki In, Kyung Ho Yoon
Fusion Science and Technology | Volume 52 | Number 4 | November 2007 | Pages 844-848
Technical Paper | First Wall, Blanket, and Shield | doi.org/10.13182/FST07-A1597
Articles are hosted by Taylor and Francis Online.
Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) blanket with Ferritic Steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The design and WKH performance of the KO HCML Test Blanket Module (TBM) are introduced in this paper. It uses He as a coolant at an inlet temperature of 300°C and an outlet temperature up to 406°C and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the Tritium Breeding Ratio (TBR) and the shielding performances. A 3-D Monte Carlo analysis is performed with the MCCARD code for the neutronics and the total TBM power is designed to be 0.675 MW at a normal heat flux from the plasma side. From the analysis results with CFX-10 for the thermal-hydraulics, the He cooling path is determined and it shows that the maximum temperature of the first wall does not exceed 550 °C at the structural materials and the coolant velocities are 50 m/sec and 25~32 m/sec at the first wall and breeding zone, respectively. The obtained temperature data is used in the thermal-mechanical analysis with ANSYS-10. The maximum von Mises equivalent stress of the first wall is 2540 MPa and the maximum deformation of it is 1.3 mm.
