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Full text (ru) Dimensioness method of assessing the conditions of thermal shock to the reactor vessel
V. I. Skalozubov1, T. V. Gablaia1, I. L. Kozlov2, E. S. Leshotnaya1
1Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, Kyiv, Ukraine
2Odesa National Polytechnic University, Odesa, Ukraine
Abstract: During the prolongation of the terms of the nuclear power plants operation the main issue is the substantiation of the possibility of the terms of the reactor vessel operation, which determines the economic necessity to continue operating the unit as a whole. Significant effect on the residual life of the reactor pressure vessel is to commit thermal loads, including thermal shock in case of the accident with leaky reactor circuit. The dimensionless method is developed concerning the rating conditions of the thermal shock to reactor vessels. Moreover, the attention is focused not on the known correlation approaches, but on the additional criteria, related to the completeness of the heat transfer and terms strength of metal under thermal loading. Proposed methods can be the basic for the rapid diagnosis of the conditions of thermal shock to reactor in the accidents with leaky reactor circuit.
Keywords: shock, thermal loads, reactor, heat, strength.
References:1. Skalozubov V. I., Klyuchnikov A. A. Basics of Extending Operation of NPPs with VVER (Chornobyl: ISP NPP NASU, 2010) 210 p. (Rus).
2. Skalozubov V. I., Klyuchnikov A. A. Scientific and Technical Basis for Improving the Safety Measures of NPPs with VVER (Chornobyl: ISP NPP NASU, 2010) 200 p. (Rus).
3. Vorob'ev Yu. Yu., Nosovskij A. V. Yadernaya i Radiatsionnaya Bezopasnost 2 (2012) (Rus).
4. IAEA. Guidance on the analysis of thermal shock for NPP with VVER. IAEA-EBR-WWER-08 (Vienna, 2008) (Rus).
5. Tuomisto H. Thermal-hydraulics of the Loviisa Reactor Pressure Vessel Overcooling transients. Imatran Voima Oy, Research report IVO-A-01/87. (1987).
6. Fluid Mixing and Flow Distribution in the Reactor Circuit (FLOMIX-R). Annual Report 2004. (Institute of Safety Research. Wissenschaftlich-technische Berichte FZR-420-2005).
7. Logvinov S. A., Bezrukov Yu. A., Dragunov Yu. G. Experimental study of thermal-hydraulic reliability of the VVER. (Podol'sk: FGUP OKB "Gidropress", 2004) (Rus).
8. International Comparative Assessment Study of Pressurized Thermal Shock in Reactor Pressure Vessels. NUREG/CR-6651 ORNL/TM-1999/231. (1999).
9. Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment Deterministic Evaluation for the Integrity of Reactor Pressure Vessel. IAEA - TECDOC - 1627 (2010).
10. Flow Stagnation and Thermal Stratification in Single and Two-Phase Natural Circulation Loops. (Italy: ICTP, 2007).
11. Kurnosov M. M., Lapatin V. M., Strebnev N. A. Thermal-hydraulic parameters of coolant mixing zones and water of ECCS in RA with VVER during leak accidents (comparison of known techniques and calculation results on them) (Podol'sk: FGUP OKB "Gidropress", 2005). (MNTK-2005). (Rus).
12. Kirillov P. L., Yur'ev Yu. S., Bobkov V. P. Handbook of Thermal-Hydraulic Calculations. 2-nd ed. (Moscow: Energoatomizdat, 1990) 360 p. (Rus).
13. Vasil'chenko V. N., Kim V. V., Skalozubov V. I. Modeling of Accidents at Nuclear Power Units of NPP (Odessa: Rezon 2000, 2002) 466 p. (Rus).