The UNICO-2F multiphysical (neutronics + thermohydraulics) code was developed for analyzing a large leak initiated in the steam generator of the BREST-OD-300 reactor by the rupturing of a steam-generating tube. The code calculates transient 3D-fields of the coolant velocity, temperature, and pressure, the concentration of the steam, the power density, and the fuel and fuel-rod cladding temperatures. The initial failure event in the analysis of core-parameter dynamics is postulated to be the rupture of a steam-generating tube. It is shown that, even under the most conservative assumptions, in the event of a full-hole rupture the reactivity introduced on steam entering the core leads to a very small power increase, the maximum temperature of the fuel rod cladding stays within acceptable limits. It is shown that even if a large leak appears simultaneously in two steam generators the power and temperature in the core are stabilized at an acceptable level. On the whole, the computational analysis confirmed the high self-protection of the reactor in relation to an accident caused by a large leak in a steam generator.
This is a preview of subscription content, log in via an institution to check access.
Similar content being viewed by others
References
Gang Wang, “A review of research progress in heat exchanger tube rupture accident of heavy liquid metal cooled reactors,” Ann. Nucl. En., 109, 1–8 (2017).
E. Bubelisa, M. Schikorr, M. Frogheri, et al., “LFR safety approach and main ELFR safety analysis results,” in: Intern. Conf. on Fast Reactor and Related Fuel Cycle: Safe Technologies and Sustainable Scenarios (FR13), IAEA, Paris, France (2013), T3-CN-199/297.
M. Frogheri, A. Alemberti, and L. Mansani, “The lead fast reactor – demonstrator (ALFRED) and ELFR design,” ibid., T2-CN-199/024.
Zhixing Gu, Gang Wang, Yunqing Bai, et al., “Preliminary investigation on the primary heat exchanger lower head rupture accident of forced circulation LBE-cooled fast reactor,” Ann. Nucl. En., 81, 84–90 (2015).
M. Jeltsov, W. Villanueva, and P. Kudinov, “Steam generator leakage in lead cooled fast reactors: modeling of void transport to the core,” Nucl. Eng. Design, 328, 255–265 (2018).
A. V. Abramov, A. N. Zyablitskikh, and A. P. Kolesnikov, “Experimental validation of the safety of the BRESTOD-300 reactor on depressurization of heat exchange tubes,” in: 3rd Int. Sci. Techn. Conf. on Innovative Nuclear Energy Projects and Technologies, NIKIET, Moscow (2014), Vol. 1, pp. 251–262.
G. B. Wallis, One-Dimensional Two-Phase Flows [Russian translation], Mir, Mosow (1972).
V. I. Berdnikov and A. M. Levin, “On the rise rate of gas bubbles in metallic and slag melts,” Izv. Vyssh Uchebn. Zaved. Chern. Metallurg., 12, 24–27 (1977).
I. R. Suslov and D. M. Babanakov, “MAG – the code for fi ne mesh VVER calculations,” in: Proc. 6th Symp. AER (1996), pp. 161–170.
Author information
Authors and Affiliations
Corresponding author
Additional information
Translated from Atomnaya Énergiya, Vol. 130, No. 4, pp. 183–188, April, 2021.
Rights and permissions
About this article
Cite this article
Rachkov, V.I., Suslov, I.R., Khomyakov, Y.S. et al. Analysis of the Consequences of a Large Leak in a Steam Generator in a Two-Circuit Lead-Cooled Reactor Unit. At Energy 130, 191–196 (2021). https://doi.org/10.1007/s10512-021-00793-w
Received:
Published:
Issue Date:
DOI: https://doi.org/10.1007/s10512-021-00793-w