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Lee, Deokjung
Computational Reactor physics & Experiment lab (CORE Lab)
Research Interests
  • Reactor Analysis computer codes development
  • Methodology development of reactor physics
  • Nuclear reactor design(SM-SFR,PWR and MSR)

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Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code

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dc.contributor.author Dos, Vutheam ko
dc.contributor.author Lee, Hyunsuk ko
dc.contributor.author Jo, Yunki ko
dc.contributor.author Lemaire, Matthieu ko
dc.contributor.author Kim, Wonkyeong ko
dc.contributor.author Choi, Sooyoung ko
dc.contributor.author Zhang, Peng ko
dc.contributor.author Lee, Deokjung ko
dc.date.available 2020-08-27T07:56:57Z -
dc.date.created 2020-08-20 ko
dc.date.issued 2020-09 ko
dc.identifier.citation NUCLEAR ENGINEERING AND TECHNOLOGY, v.52, no.9, pp.1881 - 1895 ko
dc.identifier.issn 1738-5733 ko
dc.identifier.uri https://scholarworks.unist.ac.kr/handle/201301/47845 -
dc.description.abstract The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications. (c) 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/). ko
dc.language 영어 ko
dc.publisher KOREAN NUCLEAR SOC ko
dc.title Overcoming the challenges of Monte Carlo depletion: Application to a material-testing reactor with the MCS code ko
dc.type ARTICLE ko
dc.identifier.scopusid 2-s2.0-85080072359 ko
dc.identifier.wosid 000553761000001 ko
dc.type.rims ART ko
dc.identifier.doi 10.1016/j.net.2020.02.003 ko
dc.identifier.url https://www.sciencedirect.com/science/article/pii/S1738573319304681?via%3Dihub ko
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