An international journal published by K. N. Toosi University of Technology

Document Type : Research Article


Nuclear Engineering Department‎, ‎Shahid Beheshti University‎, ‎G.C‎, ‎P.O‎. ‎Box 1983963113‎, ‎Tehran‎, ‎Iran


‎In this work‎, ‎dynamic responses of a WWER-1000 reactor in reactivity insertions are studied using a coupling method‎. ‎The ANSYS-CFX is implemented for thermal hydraulic study of the core and the point kinetic equation (PKE) is coupled as a FORTRAN subroutine‎. ‎For transient analysis of the core‎, ‎the thermal feedback of the fuel is added to coolant‎, ‎and numerical solver of cylindrical heat transfer for obtaining the irradiated fuel rod temperature profile is also included‎. ‎In order to investigate the irradiation effect‎, ‎the fuel and gap properties in burnup with appropriate correlations could be calculated‎. ‎Using memory management system (MMS) and data transfer arrays, coupling between numerical subroutines is carried out‎. ‎It is shown that the dynamic response of the core depends on burnup‎, ‎and the response could be varied in time‎. ‎In addition‎, ‎the coupling method is reliable for other dynamic calculations‎.


  • The reactivity response of an irradiated PWR core with coupling CFX and PKM is studied.
  • Eff ects of ssion gas release, dissolved gases, porosity, radiation damage and fuel Burnup are considered.
  • Using the MMS, the online data transferring from CFX and subroutines is available.
  • The response of the BNPP reactor in several reactivity insertions during burnups is studied.
  • The e ects of irradiation on dynamic response of fuel in a FA and core are shown.


Aghaie, M., Zolfaghari, A., and Minuchehr, A. (2012). Coupled neutronic thermal–hydraulic transient analysis of accidents in PWRs. Annals of Nuclear Energy, 50:158–166.

Berna, G. A., Beyer, G. A., Davis, K. L., et al. (1997). Frapcon-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup. Technical report, Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; Pacific Northwest Lab., Richland, WA (United States); Idaho National Engineering Lab., Idaho Falls, ID (United States).

Chen, Z., Chen, X.-N., Rineiski, A., et al. (2015). Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses. Annals of Nuclear Energy, 83:41–49.

FSAR (2005). Final safety assessment report (FSAR) for BNPP. Technical report, Accident Analysis, Book 4, Moscow.

Lucuta, P. G., Matzke, H. J., and Hastings, I. J. (1996). A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel: review and recommendations. Journal of Nuclear Materials, 232(2-3):166–180.

Luyben, W. L. (2012). Use of dynamic simulation for reactor safety analysis. Computers & Chemical Engineering, 40:97-109.

Olander, D. R. (1976). Fundamental aspects of nuclear reactor fuel elements: solutions to problems. Technical report, California Univ., Berkeley (USA). Dept. of Nuclear Engineering.

Smith, B. L. (2010). Assessment of CFD codes used in nuclear reactor safety simulations. Nuclear Engineering and Technology, 42(4):339–364.

Vyskocil, L. and Macek, J. (2014). Coupling CFD code with system code and neutron kinetic code. Nuclear Engineering and Design, 279:210–218.