A CFD analysis of blockage length on a partially blocked fuel rod

Authors

  • Nikolas Lymberis Scuro Instituto de Pesquisas Energéticas e Nucleares, Avenida Lineu Prestes, 2242, Centro de Engenharia Nuclear, Zip Code 05508-000, São Paulo, SP, Brazil
  • Pedro Ernesto Umbehaun Instituto de Pesquisas Energéticas e Nucleares, Avenida Lineu Prestes, 2242, Centro de Engenharia Nuclear, Zip Code 05508-000, São Paulo, SP, Brazil
  • Edvaldo Angelo Universidade Presbiteriana Mackenzie, Rua da Consolação, 930, Zip Code 01302-907, São Paulo, SP, Brazil
  • Gabriel Angelo Centro Universitário FEI, Avenida Humberto de Alencar Castelo Branco, 3972-B, Zip Code 09850-901, São Bernardo do Campo, SP, Brazil Instituto de Pesquisas Energéticas e Nucleares, Avenida Lineu Prestes, 2242, Centro de Engenharia Nuclear, Zip Code 05508-000, São Paulo, SP, Brazil
  • Delvonei Alves de Andrade Instituto de Pesquisas Energéticas e Nucleares, Avenida Lineu Prestes, 2242, Centro de Engenharia Nuclear, Zip Code 05508-000, São Paulo, SP, Brazil

DOI:

https://doi.org/10.15392/bjrs.v7i2B.437

Keywords:

Partially blocked fuel rod, CFD, ANSYS CFX

Abstract

After a loss of coolant accident (LOCA), fuel rods may balloon. The swelling can partially block the flow channel, affecting the coolability during reflood phase. In order to analyze the influence of blockage length, using a radial blockage of 90%, varying just the blockage length, many steady state numerical simulations has been done using Ansys-CFX code to verify thermal-hydraulic properties according to different forced cooled conditions. Temperature peaks are observed on cladding, followed by a temperature drop. A 5x5 fuel assembly, with 9 centered ballooned fuel rod, flow redistribution inside channels can also be captured, indicating an overheating zone. Therefore, this study conclude, for the same boundary conditions, the longer the blockage length originated after LOCA events, the higher are the clad temperatures, indicating the possibility of overheat during transient conditions on reflood.

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References

M. NAITOH, K. CHINO, AND H. OGASAWARA, Cooling mechanism during transient reflooding of a reactor fuel bundle after loss of coolant, Nucl. Eng. Des, vol. 44, no. 2, pp. 193–200, Nov. 1977.

P. IHLE AND K. RUST, “PWR reflood experiments using full length bundles of rods with zircaloy claddings and alumina pellets,” Nucl. Eng. Des., vol. 99, pp. 223–237, Feb. 1987.

M. L. ANG, A. AYTEKIN, AND A. H. FOX, “Analysis of flow distribution a PWR fuel rod bundle model containng A 90% blockage,” Nucl. Eng. Des., vol. 103, no. 2, pp. 165–188, Aug. 1987.

M. L. ANG, A. AYTEKIN, AND A. H. FOX, “Analysis of flow distribution in a PWR fuel rod bundle model containing a blockage - Part 1. A 61% coplanar blockage,” Nucl. Eng. Des., vol. 108, no. 3, pp. 275–294, Jul. 1988.

M. L. ANG, A. AYTEKIN, AND A. H. FOX, “Analysis of flow distribution in a PWR fuel rod bundle model containing a blockage - Part 2. A non-coplanar blockage,” Nucl. Eng. Des., vol. 108, no. 3, pp. 295–314, Jul. 1988.

K. R. THURGOOD, M. J.; KELLY, J. M.; GUIDOTTI, T. E.; KOHRT, R. J.; CROWELL, “COBRA/TRAC-A thermal-hydraulics code for transient analysis of nuclear reactor vessels and primary coolant systems.,” Comm., 1983.

C. GRANDJEAN, “Coolability of blocked regions in a rod bundle after ballooning under LOCA conditions,” Nucl. Eng. Des., vol. 237, no. 15–17, pp. 1872–1886, Sep. 2007.

K. IHLE, P.; RUST, “FEBA-flooding experiments with blocked arrays-influence of block-age shape,” Trans. Am. Nucl. Soc., 1979.

K. G. JOWITT, D.; COOPER, C. A., “The THETIS 80% blocked cluster experiment. Part 5,” UKAEA At. Energy Establ., vol. No. AEEW-R, 1984.

K. G. DENHAM, M. K.; JOWITT, D.; PEARSON, “ACHILLES unballooned cluster ex-periments, part 1, description of the ACHILLES rig, test section and experimental procedures.,” AEEW-R2326., 1989.

B. D. G. FAIRBAIRN, S. A. ; PIGGOTT, “Flow and Heat Transfer in PWR Rod Bundles in the Presence of Blockage due to clad Ballooning; Experimental Data Report-Part 2,” CEGB Rep. TPRD/B/0458, vol. 2, 1984.

L. E. HOCHREITER, “FLECHT SEASET program. Final report,” Westinghouse Electr. Corp., Pittsburgh, PA, vol. No. NUREG/, no. Westinghouse Electric Corp., Pittsburgh, PA (USA), 1985.

B. J. KIM, J. KIM, K. KIM, S. W. BAE, AND S.-K. MOON, “Effects of fuel relocation on reflood in a partially-blocked rod bundle,” Nucl. Eng. Des., vol. 312, pp. 239–247, 2017.

K. KIM, B.-J. KIM, H.-S. CHOI, S.-K. MOON, AND C.-H. SONG, “Effect of a blockage length on the coolability during reflood in a 2×2 rod bundle with a 90% partially blocked region,” Nucl. Eng. Des., vol. 312, pp. 248–255, 2017.

E. E. DOMINGUEZ-ONTIVEROS, Y. A. HASSAN, M. E. CONNER, AND Z. KAROUTAS, “Experimental benchmark data for PWR rod bundle with spacer-grids,” Nucl. Eng. Des., vol. 253, pp. 396–405, Dec. 2012.

E. E. DOMINGUEZ-ONTIVEROS AND Y. A. HASSAN, “Non-intrusive experimental investigation of flow behavior inside a 5×5 rod bundle with spacer grids using PIV and MIR,” Nucl. Eng. Des., vol. 239, no. 5, pp. 888–898, May 2009.

S. CHENG, H. CHEN, AND X. ZHANG, “CFD analysis of flow field in a 5×5 rod bundle with multi-grid,” Ann. Nucl. Energy, vol. 99, pp. 464–470, 2017.

H. J. WAGNER, W. ; KRETZSCHMAR, “IAPWS industrial formulation 1997 for the thermodynamic properties of water and steam,” Int. Steam Tables Prop. Water Steam Based Ind. Formul. IAPWS-IF97, p. 7–150., 2008.

B. E. LAUNDER AND D. B. SPALDING, “The numerical computation of turbulent flows,” Comput. Methods Appl. Mech. Eng., vol. 3, no. 2, pp. 269–289, Mar. 1974.

E. G. STERN, F.;WILSON, R. V.; COLEMAN, H. W.; PATERSON, “Comprehensive ap-proach to verification and validation of CFD simulations-Part 1: methodology and procedures,” Trans. Soc. Mech. Eng. J. Fluids Eng., vol. 123(4), pp. 793–802, 2001.

K. IHLE, P.; RUST, “SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3,” Kernforschungszentrum Karlsruhe GmbH, 1986.

E. DOMINGUEZ-ONTIVEROS AND Y. A. HASSAN, “Experimental study of a simpli-fied 3x3 rod bundle using DPTV,” Nucl. Eng. Des., vol. 279, pp. 50–59, Nov. 2014.

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Published

2019-06-25

Issue

Section

XX Meeting on Nuclear Reactor Physics and Thermal Hydraulics (XX ENFIR)

How to Cite

A CFD analysis of blockage length on a partially blocked fuel rod. Brazilian Journal of Radiation Sciences, Rio de Janeiro, Brazil, v. 7, n. 2B (Suppl.), 2019. DOI: 10.15392/bjrs.v7i2B.437. Disponível em: https://bjrs.org.br/revista/index.php/REVISTA/article/view/437.. Acesso em: 22 nov. 2024.

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