Development and application of a thermal-hydraulics analysis code for small modular natural circulation lead or lead-alloy cooled fast reactors

Author:

Chen ZhaoORCID,Zhou Guangming,Zhao Pengcheng,Chen Hongli

Funder

National Natural Science Foundation of China

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference15 articles.

1. Review and proposal for best fit of wire-wrapped fuel bundle friction factor and pressure drop predictions using various existing correlations;Bubelis;Nucl. Eng. Des.,2008

2. Chen, H., et al., 2012. Preliminary thermal-hydraulic design and analysis of china lead alloy cooled research reactor (CLEAR-I). The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9), 9–14 September, Kaohsiung, Taiwan.

3. Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors;Chen;Ann. Nucl. Energy,2014

4. IAEA, 2007. Status of Small Reactor Designs without On-Site Refuelling. TECDOC-1536.

5. Validation of the RELAP5 code for the modeling of flashing-induced instabilities under natural-circulation conditions using experimental data from the CIRCUS test facility;Kozmenkov;Nucl. Eng. Des.,2012

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