Analysis of the Kalinin-3 coolant transient benchmark by SKETCH-N/SKAZKA code system Chukwudi A.S. Ojinnaka, Vyacheslav G. Zimin ⇑ , Vadim P. Strashnykh, Sergei P. Nikonov Institute of Nuclear Physics and Engineering, National Research Nuclear University MEPhI, Kashirskoe Shosse, 31, Moscow 115409, Russia article info Article history: Received 16 January 2020 Received in revised form 5 May 2020 Accepted 11 July 2020 Keywords: VVER-1000 reactor Kalinin-3 coolant transient benchmark Main coolant pump trip Reactor core calculation Neutron kinetics abstract The switching off of the one of four main circulation pumps results in the Kalinin-3 Coolant Transient Benchmark. The benchmark contains four exercises, but this article shows the results of the exercise two of the VVER-1000 reactor core calculation. Reactor core analysis code SKETCH-N/SKAZKA is used to simulate the steady-state and transient conditions of the VVER reactor. The reactor core boundary con- ditions for the problem are provided by ATHLET/BIPR-VVER code. In the initial steady-state condition and at the end of the transient the power distribution is compared with the nuclear power plant (NPP) data. The average error in 3-D power distribution is 3.6%. The average difference in fuel assembly (FA) power is less than 2%. In the transient calculation, we compared with the experimental data the reactor power and axial offset. There is close agreement between the numerical results and NPP data. Ó 2020 Elsevier Ltd. All rights reserved. 1. Introduction The Kalinin NPP is currently made up of four units, where unit one and two are both of the VVER 1000/338, while the third and the fourth units are of type VVER-1000/320. The third unit - the Kalinin-3 started up on 16.12.2004. Experimental test at this unit was conducted on 02.10.2005. The experiment was the switching off of one of the four Main Circulation Pumps (MCP). This experi- ment gave rise to the Kalinin-3 Coolant Transient Benchmark (Tereshonok et al., 2008). After MCP is switched off, the control rods of group number 10 were inserted to set the power of the reactor equal to 67%. The transient results in non-symmetric power distribution. It is important to note that the benchmark contains four exercises, but this article deals only with exercise two (Tereshonok et al., 2008). Exercise two considers the transient response of the reactor core with the given boundary conditions. The reactor core calculations are performed by the SKETCH-N/ SKAZKA coupled code system, where SKETCH-N code (Zimin, 2000) performs the neutronics calculations, while the SKAZKA code (Strashnykh, 2003) simulates the reactor core thermal- hydraulics. The boundary conditions for the thermal-hydraulics core calculations are taken from the ATHLET/BIPR-VVER code (Nikonov et al., 2011). Kalinin-3 coolant transient benchmark cal- culations include the results of many codes, a few of them are: TRAP-KS and KORSAR/GP codes (Makhin et al., 2010), ATHLET/ KIKO3D code (Keresztúri et al., 2012), COBAYA3/SUBCHANFLOW code (Calleja et al., 2014), DYN3D/ATHLET (Kozmenkov et al., 2015). Generally, there are good agreements between the calcu- lated and experimental data. The main difference between the SKETCH-N/SKAZKA calcula- tions and the other code calculations, that we use our own neutron cross section library and do not use the neutron cross section library provided to the benchmark participants. We also performed the burnup calculations by the SKETCH-N/SKAZKA code from the reactor start-up till the time of the experiment. The paper is organized as follows. In the Section 2, we give the brief description of the SKETCH-N and SKAZKA codes. The descrip- tion of the Kalinin-3 Coolant Transient Benchmark is given in Section 3. Section 4 presents the boundary condition data taken from the ATHLET/BIPR-VVER code. Section 5 contains the numerical results of the steady-state and transient calculations. The conclusions of the paper are given in the Section 6. 2. Description of the SKETCH-N/SKAZKA code system To performed the steady-state, burnup and transient calcula- tions we used the coupled SKETCH-N/SKAZKA code. SKETCH-N is used for neutronics calculations, while SKAZKA code is used to simulate the reactor core thermal-hydraulics. The description of these two codes is given in this Section. https://doi.org/10.1016/j.anucene.2020.107716 0306-4549/Ó 2020 Elsevier Ltd. All rights reserved. ⇑ Corresponding author. E-mail address: vgzimin@mail.ru (V.G. Zimin). Annals of Nuclear Energy 147 (2020) 107716 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene