Analysis of experience, safety and prospects of diversification of nuclear fuel at nuclear power plants

Authors

DOI:

https://doi.org/10.15587/2312-8372.2019.180740

Keywords:

diversification of nuclear fuel, nuclear safety, reliability of the nuclear reactor core

Abstract

The object of research is the design fuel assemblies (FA-A) at nuclear power plants with water-cooled water reactors located in Ukraine. The analysis of the experience and prospects of diversification of design fuel assemblies of pressurized water reactors with alternative heat generating assemblies of Westinghouse Electric Company is carried out. The analysis is carried out on the basis of measures and results of diversification of fuel assemblies at the Temelin nuclear power plant (Czech Republic), as well as the South Ukrainian and Zaporizhzhia nuclear power plants (Ukraine). As a result of the analysis, it is shown that the diversification of design fuel assemblies by alternative heat assemblies by Westinghouse Electric Company provides the necessary nuclear safety conditions for the maximum allowable temperature of the gratings of fuel rods and the temperature of nuclear fuel.

The work determines the need for additional analysis of nuclear safety and reliability of the reactor loop equipment in relation to the conditions of critical hydrodynamic shock during the diversification of fuel assemblies. It is revealed that the known results of nuclear safety analysis during diversification of fuel assemblies by traditional accident modeling approaches are not sufficiently substantiated. And they also significantly depend on the negative effects of «differences in the results of accident modeling by different users of the same codes» and «differences in the results of accident modeling by different codes». In addition, well-known deterministic codes do not simulate the conditions and consequences of hydrodynamic shock and various types of thermohydrodynamic instability in the reactor loop. It is shown that it is necessary to develop alternative methods for analyzing nuclear safety and equipment reliability of systems important to safety in diversifying fuel assemblies that are independent of the above negative effects. A calculation analysis of the influence of the coolant speed on the external heat transfer coefficients determined that the safety conditions for the permissible temperature of the Westinghouse Electric Company fuel rods are provided up to a maximum design temperature of 90 °C in safety heat exchangers.

Author Biographies

Vladimir Skalozubov, Odessa National Polytechnic University, 1, Shevchenko ave., Odesa, Ukraine, 65044

Doctor of Technical Sciences, Professor, Laureate of the State Prize of Ukraine for Science and Technology

Intersectoral Center for Fundamental Scientific Research in the Field of Energy and Ecology

Serg Melnik, Odessa National Polytechnic University, 1, Shevchenko ave., Odesa, Ukraine, 65044

PhD, Senior Lecturer

Department of Theoretical, General and Alternative Energy

Oksana Pantak, Odessa National Polytechnic University, 1, Shevchenko ave., Odesa, Ukraine, 65044

Postgraduate Student

Department of Nuclear Power Plants

Vitalii Gryb, Odessa National Polytechnic University, 1, Shevchenko ave., Odesa, Ukraine, 65044

Postgraduate Student

Department of Nuclear Power Plants

Vladyslav Spinov, Odessa National Polytechnic University, 1, Shevchenko ave., Odesa, Ukraine, 65044

PhD, Researcher

Intersectoral Center for Fundamental Scientific Research in the Field of Energy and Ecology

Yurii Komarov, State Enterprise «National Nuclear Energy Generating Company «Energoatom», 3, Nazarivska str., Kyiv, Ukraine, 01032

Doctor of Technical Sciences, Laureate of the State Prize of Ukraine for Science and Technology

References

  1. Kirst, M., Benjaminsson, U., Önneby, C. (2015). Diversification of the VVER, Fuel Market in Eastern Europe and Ukraine. ATW, 60 (3), 63–87.
  2. Daniel, E., Milisdörfer, L. (2010). 10 years of experience with Westinghouse fuel at NPP Temelín. International Topical Meeting VVER-2010 Experience and Perspectives. Prague, 124–137.
  3. Höglund, J., Riznychenko, O., Latorre, R., Lashevych, P. (2011). Performance of the Westinghouse WWER-1000 fuel design. International conference on WWER fuel performance, modelling and experimental support. Helena Resort, 245–256.
  4. Kaichao, S. (2008). MCNP modeling of hexagon VVER fuel. Stockholm: Royal Institute of Technology, 75.
  5. Dyversyfikatsiia postachannia yadernoho palyva na ukrainski AES. Tsentr Razumkova (2009). Natsionalna bezpeka i oborona, 6, 38–49.
  6. Ivanov, V. V. (2002). Reactor core monitoring in terms of mixed fuel loading. Symposium during International Youth Nuclear festival Dysnai-2002, Technical section. Visaginas, 65–89.
  7. Shevchenko, Y. A., Vorobev, Yu. Yu. (2015). Proverka kryteryev bezopasnosty smeshannikh zahruzok yadernoho toplyva dlia reaktorov typa VVER-1000. Yaderna ta radiatsiina bezpeka, 2 (66), 3–7.
  8. Dziubenko, B. V., Ashmantas, L.-V., Segal, M. D. (1994). Modelirovanie stacionarnykh i perekhodnykh teplogidravlicheskikh processov v kanalakh slozhnoi formy. Vilnius: Pradai, 240.
  9. Grebennikov, A. N., Deulin, A. A., Manoshina, I. O. et. al. (2013). Adaptaciia, verifikaciia i ispolzovanie paketa programm LOGOS dlia resheniia zadach atomnoi energetiki. Obespechenie bezopasnosti AES s VVER. Podolsk: FGUP OKB «GIDROPRESS», 23–32.
  10. Kirillov, P. L., Iurev, Iu. S., Bobkov, V. P. (1990). Spravochnik po teplogidravlicheskim raschetam. Moscow: Energoatomizdat, 360.
  11. Bykov, M. A., Lisenkov, E. A., Bezrukov, Iu. A. et. al. (2009). Modelirovanie processov peremeshivaniia teplonositelia v reaktore kodami TRAP-KS, DKM i KORSAR/GP. Obespechenie bezopasnosti AES s VVER. Podolsk: FGUP OKB «GIDROPRESS», 45–60.
  12. Predvaritelnii otchet po obosnovaniiu bezopasnosti ispolzovaniia uprochnennoi konstrukcii TVS kompanii «Vestinkhauz» na energobloke No. 3 IUUAES (2004). Kharkiv: NNC KHFTI.CPAZ, 548.
  13. Mazurenko, A. S., Skalozubov, V. I., Chulkin, O. A., Pirkovskiy, D. S., Kozlov, I. L. (2017). Determining the Conditions for the Hydraulic Impacts Emergence at Hydraulic Systems. Problems of the Regional Energetics, 2 (34), 98–104.
  14. Skalozubov, V., Chulkin, O., Pirkovskiy, D., Kozlov, I., Komarov, Yu. (2019). Method for determination of water hammer conditions and consequences in pressurizers of nuclear reactors. Turkish Journal of Physics, 43, 229–235. doi: http://doi.org/10.3906/fiz-1809-5

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Published

2019-07-12

How to Cite

Skalozubov, V., Melnik, S., Pantak, O., Gryb, V., Spinov, V., & Komarov, Y. (2019). Analysis of experience, safety and prospects of diversification of nuclear fuel at nuclear power plants. Technology Audit and Production Reserves, 4(1(48), 26–33. https://doi.org/10.15587/2312-8372.2019.180740

Issue

Vol. 4 No. 1(48) (2019): Industrial and technology systems

Section

Technology and System of Power Supply: Original Research

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Copyright (c) 2019 Vladimir Skalozubov, Serg Melnik, Oksana Pantak, Vitalii Gryb, Vladyslav Spinov, Yurii Komarov

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