Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System

Wei Shiying; Wang Chenglong; Su Guanghui; Tian Wenxi; Qiu Suizheng

doi:10.13832/j.jnpe.2018.04.0067

Volume 39 Issue 4

Aug. 2018

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Article Navigation > Nuclear Power Engineering > 2018 > 39(4): 67-70

Wei Shiying, Wang Chenglong, Su Guanghui, Tian Wenxi, Qiu Suizheng. Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System[J]. Nuclear Power Engineering, 2018, 39(4): 67-70. doi: 10.13832/j.jnpe.2018.04.0067

Citation:

Wei Shiying, Wang Chenglong, Su Guanghui, Tian Wenxi, Qiu Suizheng. Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System[J]. Nuclear Power Engineering, 2018, 39(4): 67-70. doi: 10.13832/j.jnpe.2018.04.0067

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Development of Analysis Code for Pb-Bi Cooled Direct-Contact-Boiling Water Fast Reactor System

doi: 10.13832/j.jnpe.2018.04.0067

Publish Date: 2018-08-15

Abstract

Abstract

In this paper, mathematical models are established for PBWFR coolant system, and a system analysis code of PBWFR SACOL is developed. The steady-state and transient thermal-hydraulic performance of PBWFR is calculated and analyzed, especially the unprotected transient of over power. Results show that PBWFR is safe under steady state. However, during UTOP transient, the rapidly increasing of power in a short time would lead to cladding failure.
- Pb-Bi cooled direct-contact-boiling water fast reactor (PBWFR),
- Lead-bismuth reactor,
- Thermal-hydraulics,
- System safety