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The Molten Salt Reactor (MSR) is one of the Generation IV reactors and it provides inherent safety and high efficiency. The Fuel Drain Tank (FDT) is a critical component in MSR for storing fuel salt in case of shutdown as well as emergency and has to be designed to remove decay heat reliably for long time. To design and predict the behavior of MSR FDT, the Oak Ridge National Lab (ORNL) assumed certain parameters in 1960s while designing the Molten Salt Reactor Experiment (MSRE). One of the key assumptions is the difference between the mixed-mean temperature of the fuel salt and the thimble tube surface temperature remains constant at 25 F. This assumption greatly simplifies the heat removal prediction as well as the design of FDT. However, the assumption was not revisited for verification in the open literature. Thus, this study attempted to verify the value using the three-dimensional transient CFD simulations which were verified through the hierarchical verification and validation methodology. Consequently, it was found that the difference between FDT mixed-mean temperature and thimble tube surface-averaged temperature was maintained around 20 K, not 25 F after thermal stratification was well-developed.

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기사명 저자명 페이지 원문 목차
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Development of robust nuclear safeguards methods for gas centrifuge enrichment plants Donghyeok Koo, Junyoung Seo, Seung Min Woo, Byeongjun Kim, Kyunam Kim, Eunju Jun, Soon Heung Chang p. 1-12

Zeolite and silica fume as enhancers for cementitious solidification of borate radioactive waste Xinfeng Liu, Chunguang Li, Zhimeng Yao, Shengdong Zhang, Zhenzhong Liu, Longcheng Liu, Tianle Li p. 1-8

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Precision of moisture quantification in canisters during the vacuum desiccation process prior to the dry storage of spent nuclear fuel Ji Hwan Lim, Kyoung-Sik Bang, Seung-Hwan Yu p. 1-16

Long-term performance test of Cu canisters used for disposing of spent nuclear fuels under aerobic groundwater Junhyuk Jang, Minsoo Lee, Gha-Young Kim, Seok Yoon p. 1-9

Plain waterjet cutting for concrete plug dismantling in disposal research tunnel Yohan Cha, Hyun-Joong Hwang, Jin-Seop Kim, Tae-Min Oh p. 1-10

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Self-supervised denoising method for single neutron image based on the S2S-NR network Wangtao Yu, Peng Xu, Xinghui Cai, Man Zhou, Jie Bao p. 1-11

Design and analysis of an AB-BNCT self-shielding rotating neutron target Zijian Zhang, Zuokang Lin, Guanchao Wu,, Zhiyuan Lin, Yuchen Liu, Tianyu Ma, Hao Li, Ye Dai, Zhimin Dai p. 1-9

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Towards energy-insensitive and robust neutron/gamma classification : a learning-based frequency-domain parametric approach Pengcheng Ai, Hongtao Qin, Xiangming Sun, Kaiwen Shang p. 1-11

Size matters : radiation shielding superiority of borate glasses with nano vs. micro ZnO Aljawhara H. Almuqrin, M.I. Sayyed, M. Elsafi p. 1-8

Investigation of optical and radiation shielding properties in bismuth oxide-doped barium borate glasses Mohammad H. Alhakami, A.S. Abouhaswa, Numa A. Althubiti, Taha Abdel Mohaymen Taha p. 1-12

Organ dose estimation for occupationally exposed workers in medical practice in Gabon during the period 2013–2023 Philippe Ondo Meye, Guy-Blanchard Dallou, Beaud Conrad Mabika Ndjembidouma, Charles Chaley, Sylvère Yannick Loemba Moandza, Patrice Ele Abiama, Saidou Saidou, Germain Hubert Ben-Bolie p. 1-8

3D radiation map reconstruction strategy using Gaussian process regression with attenuation effects and radiation characteristics Woosung Cho, Sung-Gyun Shin, Min-Jae Lee, Seungsoo Jang, Janghee Lee, Sangsoo Han, HangJin Jo, Song-Hyun Kim p. 1-10

Design and validation of shielding for a neutron generator using MCNP6 Jae Chang Kim, Junehyung Lee Bernaski, Sehwan Seol, Ju Hyung Kim, Yong Kyun Kim p. 1-10

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Research of hardening strategies based on the single event effect induced by atmospheric-like neutrons Ze-Qi Huang, Han-Tao Jing, Li-Hua Mo, Zhi-Liang Hu, Gao-Peng Shen, Zhi-Hao Zhou, Chang Cai p. 1-8

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Comparative analysis of VVER-1200 core inventory considering non-uniform and uniform reactor core power distribution Md Rosaidul Mawla, Anisur Rahman, Md. Shafiqul Islam p. 1-9

Feasibility study for the criticality calculations of the LEU+ with ATF fuels Keon Il Cha, Seok Geun Cho, Kyoon Ho Cha p. 1-9

GPU-based multi-group discrete ordinates transport calculations : parallel computing implementation in STRAUM Ao Zhang, Ser Gi Hong, Seungil Jeong, Jingen Chen p. 1-12

SphiNx : a SN-DFEM neutron transport code based on the MOOSE framework Hu Zhang, Da Li, Tianliang Hu, Guangchun Zhang, Duoyu Jiang, Qianwei Chen, Henglin Hu p. 1-8

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Experimental study and sub-channel analysis on the transverse mixing of the helical cruciform fuel assembly Qi Zhang, Zhenhai Liu, Yile Zhang, Hua Pang, Mengke Cai, Zhenhan Hong, Guangliang Yang, Haoyu Wang, Li Liu, Yao Xiao, Hanyang Gu p. 1-14

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Transient CFD analysis of MSRE fuel drain tank Moon Hyeok Kang, Jeong Ik Lee p. 1-15

Experimental investigation of the effect of operating conditions on internal flow behavior in APR1000 Yo Han Kim, Kihwan Kim, Hae-Seob Choi, Hyosung Seol, Byung-Jun Lim, Woo-Shik Kim, Dong-Jin Euh p. 1-17

Numerical investigation on mixed convection in narrow rectangular channel under inclination condition Qiang Lian, Lie Wei, Luteng Zhang, Simiao Tang, Longxiang Zhu, Zaiyong Ma, Wan Sun, Liangming Pan p. 1-9

Development of advanced containment thermal-hydraulic analysis methodology based on three-field model codes for APR1400 Seong-ho Jee, Yong-Ju Cho, Kum-Ho Han, Jisu Kim, Sung-yong Kim, Je-woo Cho, Jehee Lee, Yeon-Jun Choo, Jeong-ju Kim, Chan-eok Park, Seok-jeong Park, Sun-Hong Yoon, Soon-Joon Hong p. 1-15

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