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Nuclear Technology Volume 206, 2020 - Issue 9: Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18)

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Technical Papers

Uniform Heated Scaled-Down Standard Fuel Block Test to Validate Core Thermofluid Analysis Code for Prismatic Gas-Cooled Reactor

Chan Soo KimKorea Atomic Energy Research Institute, Versatile System Technology Development Division, 111, Daedukdaero 989beon-gil, Yuseong-gu, Daejeon, Republic of KoreaCorrespondence[email protected]

https://orcid.org/0000-0002-6447-813X

Byung Ha ParkKorea Atomic Energy Research Institute, Versatile System Technology Development Division, 111, Daedukdaero 989beon-gil, Yuseong-gu, Daejeon, Republic of Korea

Eung Seon KimKorea Atomic Energy Research Institute, Versatile System Technology Development Division, 111, Daedukdaero 989beon-gil, Yuseong-gu, Daejeon, Republic of Korea

Min Hwan KimKorea Atomic Energy Research Institute, Versatile System Technology Development Division, 111, Daedukdaero 989beon-gil, Yuseong-gu, Daejeon, Republic of Korea

Pages 1397-1408 | Received 08 Dec 2019, Accepted 22 Feb 2020, Published online: 24 Apr 2020

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https://doi.org/10.1080/00295450.2020.1735228
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