A peer-reviewed journal published by K. N. Toosi University of Technology

Document Type : Research Article

Authors

1 Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran

2 Physics Department, Imam Hossein University, Tehran, Iran

Abstract

In this study, thermal-hydraulic analysis of partial loss of coolant flow accident in supercritical pressure light water reactor (SCWR) with a new geometric design has been investigated. In the new design, the coolant and moderator circuits are separated. This analysis was performed using the development of a transient-state thermal-hydraulic code in which the equations of mass, momentum, and energy are solved. The porous Media approach is used to solve these equations. By extracting the results of transition modeling, it is observed that in the new geometric design, by separating the coolant and moderator circuits, the maximum fuel clad temperature is lower than the maximum fuel clad temperature value of the previous designs. As in the new design at the end of the transition, the maximum fuel clad temperature has decreased by about 37% compared to the initial state. The result of the calculations in this study shows that the new design, in which the coolant and moderator circuits are separated, has created more safety in a chosen transition.

Highlights

• A thermal-hydraulic computer code was developed to analyze the SCWR reactor core in transient mode.
• Porous media approach for a thermal-hydraulic analysis of the SCWR core has been implemented.
• In the new design, the amount of reactor thermal power at the end of the transition reaches 5.5% of the initial value.
• The results have been improved compared with those of Oka’s design.

Keywords

Bahrevar, M., Jahanfarnia, G., Shayesteh, M., et al. (2018). Thermal-hydraulic analysis of a novel design super critical water reactor with Al2O3 nanofluid as a coolant. The Journal of Supercritical Fluids, 140:41–52.
Broeders, C., Sanchez, V., Stein, E., et al. (2003). Validation of coupled neutron physics and thermal-hydraulics analysis for HPLWR. ICAPP03.
Buongiorno, J. and MacDonald, P. (2003). Supercritical water reactor (SCWR). Progress Report for the FY-03 Generation- IV R&D Activities for the Development of the SCWR in the US, INEEL/Ext-03-03-01210, INEEL, USA, September.
Cheng, X., Schulenberg, T., Bittermann, D., et al. (2003). Design analysis of core assemblies for supercritical pressure conditions. Nuclear Engineering and Design, 223(3):279–294.
Dobashi, K., Oka, Y., and Koshizuka, S. (1997). Core and plant design of the power reactor cooled and moderated by supercritical light water with single tube water rods. Annals of Nuclear Energy, 24(16):1281–1300.
IAEA (2002). A Technology Roadmap for Generation IV Nuclear Energy Systems. https://www.gen-4.org/gif/jcms/c_40481/technology-roadmap.
Jahanfarnia, G., Tashakor, S., Salehi, A., et al. (2013). Variable moderation high performance light water reactor (VMHWR). Annals of Nuclear Energy, 58:1–5.
Licht, J., Anderson, M., and Corradini, M. (2008). Heat transfer to water at supercritical pressures in a circular and square annular flow geometry. International Journal of Heat and Fluid Flow, 29(1):156–166.
Liu, X., Yang, T., and Cheng, X. (2013a). Development and assessment of a sub-channel code applicable for trans-critical transient of scwr. Nuclear Engineering and Design, 262:499–509.
Liu, X., Yang, T., and Cheng, X. (2013b). Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code. Annals of Nuclear Energy, 59:194–203.
Oka, Y., Koshizuka, S., Ishiwatari, Y., et al. (2010). Super light water reactors and super fast reactors: super critical pressure light water cooled reactors. Springer Science & Business Media.
Oka, Y., Koshizuka, S., and Yamasaki, T. (1992). Direct cycle light water reactor operating at supercritical pressure. Journal of Nuclear Science and Technology, 29(6):585–588.
Okano, Y., Koshizuka, S.-i., and Oka, Y. (1994). Design of water rod cores of a direct cycle supercritical-pressure light water reactor. Annals of Nuclear Energy, 21(10):601–611.
Todreas, N. E. and Kazimi, M. S. (2001). Nuclear systems: elements of thermal hydraulic design. Taylor & Francis.
Wang, L., Zhao, W., Chen, B., et al. (2015). Development of three dimensional transient analysis code STTA for SCWR core. Annals of Nuclear Energy, 78:26–32.
Zhu, D., Tian, W., Zhao, H., et al. (2013). Comparative study of transient thermal–hydraulic characteristics of SCWRs with different core design. Annals of Nuclear Energy, 51:135–145.