Verification of the reactor core sub-channel thermal hydraulic analysis code: CORTH under the flow transient conditions-Reference-Cited by-同舟云学术

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Verification of the reactor core sub-channel thermal hydraulic analysis code: CORTH under the flow transient conditions

Published:2025-09 Issue: Volume:220 Page:111538
ISSN:0306-4549
Container-title:Annals of Nuclear Energy
language:en
Short-container-title:Annals of Nuclear Energy

Author:

Wang Xiaoyu,Peng Shinian,Deng Jian,Liu Yu,Tan Sichao,Wang Jiageng,Zeng Wei,Zhang Xue

Funder

National Natural Science Foundation of China

Publisher

Elsevier BV

Reference21 articles.

1. Cross Flow Mixing between Parallel Flow Channels during Boiling, Part 1;Rowe,1967

2. Rowe, D.S., 1970. COBRA-II: a Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements. Pacific Northwest Laboratories.

3. COBRA-III: a Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements;Rowe;Pacific Northwest Laboratories,1971

4. COBRA-IIIC: a Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements;Rowe;Pacific Northwest Laboratories,1973

5. VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores;Rowe,1983

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