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MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water

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Abstract

Presented here is the MIF-SKD computation code for channel-by-channel thermohydraulic calculations of fuel-rod assemblies cooled by supercritical-pressure water. The code allows one to calculate distributions of temperature and velocity of the coolant in channels of fuel-rod assemblies and temperatures of fuel-rod claddings and fuel-rod assembly casings with regular and irregular geometrical characteristics at various distributions of energy release along the length and the radius of fuel-rod assemblies.

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References

  1. Y. Oka and S. Koshizuka, “Design Concept of Once-Through Cycle Supercritical Pressure Light Water Cooled Reactors,” in Proc. of the First Int. Symposium on Supercritical Water-Cooled Reactors, Design and Technology. Tokyo, Japan. 2000. Rep. 101. P. 1–22.

  2. G. Heusener, M. Muller, T. Schulenberg, and D. Squarer, “A European Development Program for a High Performance Light Water Reactor (HPLWR),” in Proc. of the First Int. Symposium on Supercritical Water-Cooled Reactors, Design and Technology. Tokyo, Japan. 2000. Rep. 102. P. 23–28.

  3. S. J. Bushby, G. R. Dimmick, R. B. Duffey, et al., “Conceptual Design for Advanced, High-Temperature CANDU Reactors,” in Proc. of the First Int. Symposium on Supercritical Water-Cooled Reactors, Design and Technology. Tokyo, Japan. 2000. Rep. 103. P. 29–36.

  4. Y. Oka, “Review High Temperature Water and Steam Cooled Reactor Concept,” in Proc. of the First Int. Symposium on Supercritical Water-Cooled Reactors, Design and Technology. Tokyo, Japan. 2000. Rep. 104. P. 37–57.

  5. V. A. Grabezhnaya and P. L. Kirillov, “On Calculations of Heat Transfer in Tubes and Rod Bundles under Supercritical-Pressure Water Flow,” Review FEI-0297, (TsNIIatominform, Moscow, 2003).

    Google Scholar 

  6. RTM 1604.008-88. Methodological Guidelines and Recommendations on Thermohydraulic Calculations of Fast Reactor Cores (FEI, 1988).

  7. Yu. D. Baranaev, P. L. Kirillov, V. M. Poplavskii, and V. N. Sharapov, “Nuclear Reactors Using Supercritical-Pressure Water,” Atomnaya Energiya, 96(5), 374–380 (2004).

    Google Scholar 

  8. A. A. Aleksandrov, “IAPWS-IF97 System of Equations Used for Calculating Thermo-dynamic Properties of Water and Steam in Industrial Designs,” Teploenergetika, Part 1: No. 9, 69–77 (1998); Part 2, No. 10, 64–72 (1998).

  9. P. A. Ushakov, A. V. Zhukov, and P. A. Titov, “Generalization of Experimental Data on Heat Transfer to Water in Staggered Rod Bundles,” Preprint FEI-526 (Obninsk, 1974).

  10. A. V. Zhukov, N. M. Matyukhin, and Ye. Ya. Sviridenko, “Effect of Lattice Deformation on Temperature Fields and Heat Transfer of Fuel Rods of Fast Reactor,” Preprint FEI-979 (Obninsk, 1980).

  11. B. V. Duadyakin and A. S. Popov, “Heat Transfer and Pressure Drop of Tightly Packed Seven-Rod Bundle Cooled by Water Flow of Supercritical Parameters,” Trudy VTI No. 11, 244–253 (1977)

  12. P. L. Kirillov, A. A. Kolosov, and E. A. Petrova, et al. “Distribution of Temperature in Turbulent Water Flow of Supercritical Pressures (Round Tube),” Preprint FEI-1766 (Obninsk, 1986).

  13. P. L. Kirillov, V. V. Lozhkin, and A. M. Smirnov, “Investigation of Boundaries of Deteriorated Modes of a Channel at Supercritical Pressure,” Preprint FEI-2988 (Obninsk, 2003).

  14. G. Marcoczy, “Konvektive Warmeubertragung in Langsangestromten Stabbundeln bei Turbulenter Stromung, Warme — und Stoffubertrag, 5(4), 204–212 (1972)

    Google Scholar 

  15. P. L. Kirillov, A. N. Opanasenko, R. S. Pomet’ko, and A. S. Shelegov, “Experimental Investigation of Heat Transfer in Fuel Rod Bundle at Supercritical Parameters of Freon-12,” Preprint FEI-3075 (Obninsk, 2006).

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Authors and Affiliations

  1. Institute for Physics and Power Engineering (GNTs RF-FEI), State Scientific Centre of the Russian Federation, pl. Bondarenko 1, Obninsk, Kaluga oblast, 249033, Russia

    G. P. Bogoslovskaya, A. A. Karpenko, P. L. Kirillov & A. P. Sorokin

Authors
  1. G. P. Bogoslovskaya
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  2. A. A. Karpenko
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  3. P. L. Kirillov
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  4. A. P. Sorokin
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Additional information

Original Russian Text © G.P. Bogoslovskaya, A.A. Karpenko, P.L. Kirillov, A.P. Sorokin, 2009, published in Teploenergetika.

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Bogoslovskaya, G.P., Karpenko, A.A., Kirillov, P.L. et al. MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water. Therm. Eng. 56, 214–217 (2009). https://doi.org/10.1134/S0040601509030057

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  • DOI: https://doi.org/10.1134/S0040601509030057

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