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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.”
Y. F. Chen, Y. F. Chiou, S. J. Chang, S. H. Jiang, R. J. Sheu
Nuclear Technology | Volume 182 | Number 2 | May 2013 | Pages 224-234
Regular Technical Paper | Special Issue on the Symposium on Radiation Effects in Ceramic Oxide and Novel LWR Fuels / Radiation Transport and Protection | doi.org/10.13182/NT13-A16432
Articles are hosted by Taylor and Francis Online.
Surface dose rate distribution over a spent nuclear fuel dry storage cask was realistically evaluated using the MONACO with Automated Variance Reduction using Importance Calculations (MAVRIC) computational sequence in the SCALE6 code system, with special emphasis on the effects of detailed modeling on the source term and cask geometry. The first storage cask in Taiwan has been fabricated and will be ready for loading of the designated spent fuels from Taiwan Power Company's first nuclear power plant. A test run is scheduled for 2013.Neutron and gamma-ray source terms of the first batch of 56 spent fuels were determined one by one according to their specifications, burnup histories, and cooling times. The geometry of the cask was modeled in detail including the prescribed loading pattern of 56 spent fuels in the canister. MAVRIC was modified to allow specification of the source intensity and the axial distribution for each fuel bundle, and this resulted in a factor of 3 difference in the calculated surface dose rates from fuel gammas. The main purpose for such comprehensive and detailed modeling was to compare the results with a simplified model and to predict a dose rate distribution as realistically as possible in preparation for making a high-quality comparison with field measurements. In addition to checking assumptions adopted in the safety analysis report, the results of this study can provide useful guidance for the preparation of a health physics program during the test run and, more importantly, pave the way for establishing a valuable benchmark problem.
