PSI - Issue 54

Oleksii Ishchenko et al. / Procedia Structural Integrity 54 (2024) 241–249 Yaroslav Dubyk et al./ Structural Integrity Procedia 00 (2023) 000 – 000

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residual deformation, to ensure that coolability criteria for PVI are satisfied. The approximate elastic model was used for FAD structural integrity estimation, which combines fracture and plastic collapse failure modes. The results shows that safe operation of CB during LB LOCA is ensured. References API 579-1/ASME FFS-1 Fitness-For-Service, June 2016. Bader, H., Schygulla, U., Scholl, K.-H., Häfner, W., Fischer, K., Schall, M., Wolf, L., 1985. Comprehensive assessment of experimental and analytical large LOCA induced core barrel dynamics and stresses, Transactions of the 8 International Conference on Structural Mechanics in Reactor Technology SMiRT-8. Amsterdam, Netherlands, F1 1/1*. Dubyk, Y., Filonov, V., Ishchenko, O., Orynyak, I., Filonova, Y., 2018. Dynamic assessment of the Core Barrel during Loss of Coolant Accident, Proceedings of the 2018 ASME Pressure Vessels and Piping Conference. Prague, Czech Republic, PVP2018-84762 Dubyk, Y., Ishchenko, O., Kryshchuk, M., 2020. A new simple method for shell vibration analysis with initial stress accounting. Procedia Structural Integrity 26, 422 – 429. Gal, P., Svrcek, M., Pistora, V., 2019. Fluid Structure Interaction Method In Assessment Of Dynamic Response Of VVER 440 Reactor Internals To Pressure Shock Induced By Large LOCA Accident, Transactions of the 25 International Conference on Structural Mechanics in Reactor Technology SMiRT-25. Charlotte, NC, USA. VERLIFE 2013Guidelines for Integrity and Lifetime Assessment of Components and Piping in VVER Nuclear Power Plants (IAEA-NULIFE VERLIFE), Appendix C Hermansky P, Krajcovi M. 2011. The numerical simulation of the WWER440/V213 reactor pressure vessel internals response to maximum hypothetical Large-break Loss of Coolant Accident, Nuclear Engineering and Design V.241 1177–1183 Ishchenko, O., Filonov, V., Dubyk, Y., 2021. Dynamic Model of the VVER-1000 Reactor for Seismic and LB LOCA Evaluation, Proceedings of the 28th International Conference on Nuclear Engineering. Volume 4: Student Paper Competition. Virtual, Online, V004T14A085. Lee S-J, Lee E., Lee C., Park N-C., Choi Y., On C 2021. Investigation of seismic responses of reactor vessel and internals for beyond-design basis earthquake using elasto-plastic time history analysis, Nuclear Engineering and Technology V.53, 988-1003 Lund, H., Flåtten, T., 2010. Equilibrium conditions and sound velocities in two-phase flows, presented at SIAM Annual Meeting. Pittsburgh, Pennsylvania, USA. Pistora, V., Gal, P., Svrcek, M., 2019. Dynamical assessment of VVER reactor internals for large break LOCA, Transactions of the 25 International Conference on Structural Mechanics in Reactor Technology SMiRT-25. Charlotte, NC, USA. PNAE G-7-002-86. Strength calculations norms for nuclear power plant equipment and piping. – Moscow: Energoatomizdat, 1989. (in Russian). Volkov, V.Yu., Golibrodo, L.A., Krutikov, A.A., Kudryavtsev, O.V., Nadinsky, Yu.N., Nechaev, A.T., Skibin, A.P., 2017. Multiscale problems of heat and mass transfer in nuclear energy, Bulletin of the South Ural State University. Series: Computational Mathematics and Informatics. 2017. V. 6, No. 4. 60–73. (in Russian). Wang, M., Wang, L., Jing, Ji., Gao, X., Qiu, S., Tian, W., Su, G.H., 2017. Study on the reactor core barrel instantaneous characteristics in case of Loss of Coolant Accident (LOCA) scenarios for loop-type PWR. Nuclear Engineering and Design 324, 93–102. Zeman, V., Hlaváč, Z., 2018. Modelling of the friction -vibration interactions in reactor core barrel couplings. Applied and Computational Mechanics 12, 193–212. Zhang, Yu, He, Ch., Sun, L., 2021. Mathematical model for predicting the vibration performance of a core barrel considering the interaction of seismic load and fluid force. Nuclear Engineering and Design 383, 1–11. Zhang, Yu, Li, P., 2023. Transient response of fluid-surrounded core barrel under earthquake excitations bearing different frequency contents. Nuclear Engineering and Design 407, 112300.