A Practical Method for Whole-Core Thermal Analysis of a Prismatic Gas-Cooled Reactor-Reference-Cited by-同舟云学术

A Practical Method for Whole-Core Thermal Analysis of a Prismatic Gas-Cooled Reactor

Published:2012-03 Issue:3 Volume:177 Page:352-365
ISSN:0029-5450
Container-title:Nuclear Technology
language:en
Short-container-title:Nuclear Technology

Author:

Tak Nam-Il¹,Kim Min-Hwan¹,Lim Hong Sik¹,Noh Jae Man¹

Affiliation:

1. Korea Atomic Energy Research Institute, 1045 Daedeok Street Yuseong-gu, Daejeon 305-353, Korea

Publisher

Informa UK Limited

Subject

Condensed Matter Physics,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Link

https://www.tandfonline.com/doi/pdf/10.13182/NT12-A13480

Reference15 articles.

1. “A Technology Roadmap for Generation IV Nuclear Energy Systems,” GIF-002-00, U.S. Department of Energy Nuclear Energy Research Advisory Committee and Generation IV International Forum (2002).

2. “Summary for the Next Generation Nuclear Plant Project in Review,” INL/EXT-10-19142, Rev.1, Idaho National Laboratory (2010).

3. A STUDY OF A NUCLEAR HYDROGEN PRODUCTION DEMONSTRATION PLANT

4. PERSPECTIVES OF NUCLEAR HEAT AND HYDROGEN

5. THE BENCHMARK CALCULATIONS OF THE GAMMA+ CODE WITH THE HTR-10 SAFETY DEMONSTRATION EXPERIMENTS

Cited by 7 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. ONE-DIMENSIONAL NETWORK MODELLING OF A VHTR PRISMATIC BLOCK FUEL ASSEMBLY;Proceeding of International Heat Transfer Conference 17;2023

2. DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE;EPJ Web of Conferences;2021

3. CFD and systems thermal-hydraulic analysis in the design and safety assessment of high-temperature reactors;Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment;2019

4. Investigation and Assessment of the CFD for Horizontal Flow in the VHTR Core;Science and Technology of Nuclear Installations;2017

5. Fuel temperature prediction using a variable bypass gap size in the prismatic VHTR;Nuclear Engineering and Design;2016-04