Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Calculation of the Neutronic Parameters in Fast Nuclear Reactors by Using Monte Carlo Method

Document Type : Research Paper

Abstract
In this study, a computer program is implemented to calculate the neutronic parameters of a multiplier system by Monte Carlo method. This program is able to perform the calculation of various parameters such as the effective multiplication factor, neutron flux distribution, and effective delayed neutrons of the system. This program calculates the prompt and the total multiplication factor of neutrons separately, then it can be used to calculate the effective fraction of delayed neutrons by the use of the eigenvalues and also the prompt method. The results obtained for the effective neutron multiplication factor and the neutron flux distribution are compared with the experimental measurements and the results of using MCNP4C code. In this approach a good agreement between them was obtained. The comparison between the obtained results for the effective fraction of delayed neutrons of the prompt method with those of the experimental measuremants and other applied methods showed that the prompt method is a suitable approach for the calculation of the effective fraction of delayed neutrons.

Highlights

  1. W. W. Engle, Multigroup one-dimensional discrete ordinates transport code system with anisotropic scattering, Oak Ridge National Laboratory (1973).

 2.   W. A. Rhoades, D. B. Simpson, R. L. Childs, The DOT-IV two-dimensional discrete ordinates transport code with space-dependent mesh and quadrature, Oak Ridge National Laboratory (1979).

 3.   W. A. Rhoades and D. B. Simpson, The TORT three-dimensional discrete ordinates neutron/photon transport code, ORNL/TM-13221 (1997).

 4.   J. F. Briesmeister, MCNP-A general Monte Carlo N-particle transport code, Version 4C, LA-13709-M. Los Alamos National Laboratory, USA (2000).

 5.   J. J. Duderstadt and L. J. Hamilton, Nucler reactor analysis, John Wiley & Sons, Inc (1976) 61-65.

 6.   G. R. Keepin, Physic of nuclear kinetics, Aadision-Wesley Publishing Company, inc., 73-129 (1965) 161-168.

 7.   R. Klein Meulekamp and S. C. Van Der Marck, Calculating the effective delayed neutron fraction with Monte Carlo, Nuclear Science and Engineering, 152 (2006) 142-148.

 8.   G. D. Spriggs and R. D. Bush, J. M. Campbell, Calculation of the delayed neutron effectiveness factor using ratio of k-eigenvalus, Annals of Nuclear Energy, 28 (2001) 477-487.

 9.   M. Shayesteh, M. Shahriari, G. Raisali, Simulation of time dependent neutron transport in fission reactors using Monte-Carlo method, Journal of Nuclear Science and Technology, 39 (2007) 1-8.

 10. M. Shayesteh, M. Shahriari, Calculation of time-dependent neutronic parameters using Monte Carlo method, Annals of Nuclear Energy, 36 (2009) 901-909.

 11. S. A. H. Feghhi, M. Shahriari, H. Afraideh, Calculation of neutron importance function in fissionable assemblies using Monte Carlo method, Annals of Nuclear Energy, 34 (2008) 514-520.

 12. R. E. Peterson and G. A. Newby, An unreflected U-235 critical assembly, Nuclear Science and Engineering, (1956) 1-112.

 13. H. C. Paxton, Fast critical experiments, Progress in Nuclear Energy, 7 (1981) 151-174.

 14. G. E. Hansen and W. H. Roach, Six and sixteen group cross sections for fast and intermediate critical assemblies, LAMS-2543, Los Alamos Scientific Laboratory (1961).

 15. ENDF-6 Formats Manual, National nuclear data center, Brookhaven National Laboratory (2005).

 16. R. E. MacFarlane, D. W. Muir, The NJOY nuclear data processing system version91, Lose Alamos National Laboratory (1994).

 17. R. D. Mosteller, S. C. Frankle, P. G. Young, Data testing of ENDF/B-VI with MCNP: critical experiments, thermal-reactor lattices, and time-of-flight measurements, Lose Alamos National Laboratory (1992).

Keywords

Effective Delayed Neutron Fraction
Effective Multiplication Factor
Neutron Flux
Monte Carlo Method
  1. W. W. Engle, Multigroup one-dimensional discrete ordinates transport code system with anisotropic scattering, Oak Ridge National Laboratory (1973).

 2.   W. A. Rhoades, D. B. Simpson, R. L. Childs, The DOT-IV two-dimensional discrete ordinates transport code with space-dependent mesh and quadrature, Oak Ridge National Laboratory (1979).

 3.   W. A. Rhoades and D. B. Simpson, The TORT three-dimensional discrete ordinates neutron/photon transport code, ORNL/TM-13221 (1997).

 4.   J. F. Briesmeister, MCNP-A general Monte Carlo N-particle transport code, Version 4C, LA-13709-M. Los Alamos National Laboratory, USA (2000).

 5.   J. J. Duderstadt and L. J. Hamilton, Nucler reactor analysis, John Wiley & Sons, Inc (1976) 61-65.

 6.   G. R. Keepin, Physic of nuclear kinetics, Aadision-Wesley Publishing Company, inc., 73-129 (1965) 161-168.

 7.   R. Klein Meulekamp and S. C. Van Der Marck, Calculating the effective delayed neutron fraction with Monte Carlo, Nuclear Science and Engineering, 152 (2006) 142-148.

 8.   G. D. Spriggs and R. D. Bush, J. M. Campbell, Calculation of the delayed neutron effectiveness factor using ratio of k-eigenvalus, Annals of Nuclear Energy, 28 (2001) 477-487.

 9.   M. Shayesteh, M. Shahriari, G. Raisali, Simulation of time dependent neutron transport in fission reactors using Monte-Carlo method, Journal of Nuclear Science and Technology, 39 (2007) 1-8.

 10. M. Shayesteh, M. Shahriari, Calculation of time-dependent neutronic parameters using Monte Carlo method, Annals of Nuclear Energy, 36 (2009) 901-909.

 11. S. A. H. Feghhi, M. Shahriari, H. Afraideh, Calculation of neutron importance function in fissionable assemblies using Monte Carlo method, Annals of Nuclear Energy, 34 (2008) 514-520.

 12. R. E. Peterson and G. A. Newby, An unreflected U-235 critical assembly, Nuclear Science and Engineering, (1956) 1-112.

 13. H. C. Paxton, Fast critical experiments, Progress in Nuclear Energy, 7 (1981) 151-174.

 14. G. E. Hansen and W. H. Roach, Six and sixteen group cross sections for fast and intermediate critical assemblies, LAMS-2543, Los Alamos Scientific Laboratory (1961).

 15. ENDF-6 Formats Manual, National nuclear data center, Brookhaven National Laboratory (2005).

 16. R. E. MacFarlane, D. W. Muir, The NJOY nuclear data processing system version91, Lose Alamos National Laboratory (1994).

 17. R. D. Mosteller, S. C. Frankle, P. G. Young, Data testing of ENDF/B-VI with MCNP: critical experiments, thermal-reactor lattices, and time-of-flight measurements, Lose Alamos National Laboratory (1992).

Journal of Nuclear Science, Engineering and Technology (JONSAT)
Volume 34, Issue 2 - Serial Number 64
September 2012
Pages 28-37

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