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2026 Annual Conference
May 31–June 3, 2026
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My Story: John L. Swanson—ANS member since 1978
. . . and in 2019, on his 90th birthday.
Swanson in 1951, the year of his college graduation . . .
My pre-college years were spent in a rural suburb of Tacoma, Wash. In 1947, I enrolled in Reed College, a small liberal arts school in Portland, Ore.; I majored in chemistry and graduated in 1951. While at Reed, I met and married a young lady with whom I would raise 3 children and spend the next 68 years of my life—almost all of them in Richland, Wash., where I still live.
I was fortunate to have a job each of my “college summers” that provided enough money to cover my college costs for the next year; I don’t think that is possible these days. My job was in the kitchen/dining hall of a salmon cannery in Alaska. Room and board were provided and the cannery was in an isolated location, so I could save almost every dollar of my salary.
Jamal Al Zain, O. El Hajjaji, T. El Bardouni, Y. Boulaich
Nuclear Science and Engineering | Volume 193 | Number 11 | November 2019 | Pages 1276-1289
Technical Paper | doi.org/10.1080/00295639.2019.1622927
Articles are hosted by Taylor and Francis Online.
This study aims to evaluate a simplified one-dimensional thermal-hydraulic module (THM) established by the DRAGON5/DONJON5 codes that allow a multiphysics study of the Syrian Miniature Neutron Source Research (MNSR) reactor both in steady-state and transient conditions. The purpose of this paper is therefore to describe the THM, fully integrated and implanted in DONJON5 to allow coupling with neutronic modules existing in the same code and to perform steady-state thermal-hydraulic and safety analyses of the reactor. Then we compare the results given by the THM with the results obtained by the Program for the Analysis of REactor Transients (PARET)/Argonne National Laboratory (ANL) thermal-hydraulic code. In order to validate our PARET/ANL and the THM in DONJON5, the fuel center temperature as a function of core power was calculated and compared with the corresponding values of the PARET code. Moreover, we have calculated the departure from nucleate boiling ratio. The comparison of the results of this study showed a good correlation between the values obtained with the THM and the thermal-hydraulic PARET/ANL code.
