Development and Validation of Cold-Wall Correction for Correlations to Determine Critical Heat Flow in Fuel Assemblies of VVER Type Reactors

9/17/2026 2026 - #03 Thermal physics and thermal hydraulics

Ovsyannikov D.E. Vertikov E.A. Oleksyuk D.A. Zubkov A.G.

https://doi.org/10.26583/npe.2026.3.07

The article presents the results of a study of the effect of cold-wall located in the fuel assembly of a pressurized water reactor on the critical heat flux. A physical process is described that leads to an uneven distribution of local coolant parameters in a subchannel with cold-wall. Based on the physical model of dividing a subchannel into “hot” and “cold” regions, a correction factor for correlations for the definition of critical heat flux has been developed. Six experimental fuel assembly models were selected for calculational analysis, three of which contained a guide tube simulator (GT) in place of the central rod. Using a program for subchannel calculating SC-INT, in which a correction factor was introduced into the code, 454 experimental modes were calculated with critical heat flux fixation, 310 of which were obtained at the thermophysical stand KS of the NRC “Kurchatov Institute”. The correction factor is validated for fuel assemblies with triangular packing of rods. Quantitative estimates of the decrease of critical heat flux in a subchannel with cold-wall are given, and the dependence of the correction factor on the local parameters of the coolant is demonstrated.

References

1. Klemin A.I., Polyanin L.N., Strigulin M.M. Thermohydraulic calculation and thermal reliability of nuclear reactors. Moscow, Atomizdat, 1980, 261 p.
2. Rules of nuclear safety of reactor installations of nuclear power plants: NP-082-07: approved by resolution of Rostechnadzor of the Russian Federation dated 10.12.2007. Moscow, 2008.
3. Li T., Yang B.-wen, Lyu H. Cold wall effects of control rod guide tubes and experimental flow channel walls. Nuclear Engineering and Design. 2020;356:110325. DOI: https://doi.org/10.1016/j.nucengdes.2019.110325
4. Schraub F.A., Simpson R.L., Janssen E. Two-Phase Flow and Heat Transfer in Multirod Geometries: Air-Water Flow Structure Data for a Round Tube, Concentric and Eccentric Annulus, and Nine-Rod Bundle. Schenectady, NY, General Electric Company, 1969, GEAP-5739.
5. Chieng C.C., Lin C. Velocity distribution in the peripheral subchannels of the CANDU-type 19-rod bundle. Nuclear Engineering and Design. 1979;55(3):389–394. DOI: https://doi.org/10.1016/0029-5493(79)90117-1
6. Kornienko Yu.N. Distribution parameters and form factors in quasi-one-dimensional modeling of two-phase nonequilibrium flows. Thermal power engineering. 2004;7:53–63. URL: https://www.elibrary.ru/download/elibrary_27533237_92173287.pdf (accessed 05.05.2026) (in Russian).
7. Lee C.H., Mudawwar I.A. A Mechanistic Critical Heat Flux Model for Subcooled Flow Boiling Based on Local Bulk Flow Conditions. International Journal of Multiphase Flow. 1988;14(6):711–725. DOI: https://doi.org/10.1016/0301-9322(88)90070-5
8. Dmitriev S.M., Lukyanov V.E., Samoilov O.B. Substantiation of correlation for calculating critical heat flow in alternative design heat generating assemblies with mixing grating intensifiers for VVER-1000. Izvestiya vuzov. Yadernaya Energetika. 2012;1:99–108. URL: https://static.nuclear-power-engineering.ru/articles/2012/01/12.pdf (accessed Jul. 31, 2026) (in Russian).
9. Tong L.S. Boiling heat transfer and two-phase flow. Routledge, 2017, 572 p. DOI: https://doi.org/10.1201/9781315138510
10. Tong L.S. An evaluation of the departure from nucleate boiling in bundles of reactor fuel rods. Nuclear Science and Engineering. 1968;33(1):7–15. DOI: https://doi.org/10.13182/NSE68-A20912
11. Tong L.S. Prediction of departure from nucleate boiling for an axially non-uniform heat flux distribution. Journal of Nuclear Energy. 1967;21(3):241–248. DOI: https://doi.org/10.1016/s0022-3107(67)90054-8
12. Wen-Shan Lin, Bau-Shei Pei, Chien-Hsiung Lee. Bundle Critical Power Predictions under Normal and Abnormal Conditions in Pressurized Water Reactors. Nuclear Technology. 1992;98(3):354–365. DOI: https://doi.org/10.13182/NT92-A34665
13. Fighetti C.F., Reddy D.G. Parametric Study of CHF Data. Vol. 3: Critical Heat Flux Data. Palo Alto, CA, Electric Power Research Institute. 1983, EPRI NP-2609.
14. Dae-Hyun Hwang, Se-Young Chun, Keung-Koo Kim, Chung-Chan Lee. Mass velocity and cold-wall effects on critical heat flux in an advanced light water reactor. Nuclear Engineering and Design. 2007;237:369–376. DOI: https://doi.org/10.1016/j.nucengdes.2006.04.034
15. Vertikov E.A., Zporzhin K.V., Oleksyuk D.A., Khamaza V.A., Khudykin A.M., Glazov M.A., Morozkin O.N. Modernization of Subchannel Thermal-Hydraulic Code SC-INT. Izvestiya vuzov. Yadernaya Energetika. 2025;4:29–43. DOI: https://doi.org/10.26583/npe.2025.4.03 (in Russian).
16. Vertikov E.A., Oleksyuk D.A., Zubkov A.G., Malyutin M.A. On the issue of validation of core codes for calculating active zones of VVER type reactors. Problems of atomic science and technology. Series: Nuclear and reactor constants. 2025;1:232–244. EDN: MSMICP (in Russian).
17. Bezrukov Yu.A., Astakhov V.I., Saliy L.A., Lotvinov S.A., Brantov V.G., Testov I.N., Stekolnikov V.V. Investigation of critical heat fluxes in rod bundles for VVER reactors. In Proceedings of the Thermal Physics Seminar of CMEA Countries TF-74. Moscow, 1974, p. 57–66. URL: https://inis.iaea.org/records/20n1h-rny46/files/9386674.pdf?download=1 (accessed May 05, 2026) (in Russian).

departure from nucleate boiling critical heat flux cold-wall guide tube subchannel analysis experimental fuel assembly models cold-wall correction factor DNBR VVER

Link for citing the article: Ovsyannikov D.E., Vertikov E.A., Oleksyuk D.A., Zubkov A.G. Development and Validation of Cold-Wall Correction for Correlations to Determine Critical Heat Flow in Fuel Assemblies of VVER Type Reactors. Izvestiya vuzov. Yadernaya Energetika. 2026, no. 3, pp. 95-110; DOI: https://doi.org/10.26583/npe.2026.3.07 (in Russian).