Improvements and validation of the transient analysis code MOREL for molten salt reactors. Issue 8 (3rd August 2017)

• Record Type: Journal Article
• Authors: Zhuang, Kun; Zheng, Youqi; Cao, Liangzhi; Hu, Tianliang; Wu, Hongchun
• Is Part Of: Journal of nuclear science and technology. Volume 54:Issue 8(2017)
• Page Start: 878
• Page End: 890
• Publication Date: 2017-08-03
• Subjects: Molten salt reactor -- neutronics -- thermal hydraulics -- dynamics; Nuclear engineering -- Periodicals; Nuclear physics -- Periodicals; Nuclear energy -- Periodicals; 621.4805
• DOI: 10.1080/00223131.2017.1323689
• Languages: English
• ISSNs: 0022-3131

Abstract

ABSTRACT: The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.