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

Calculation of thermal neutron diffusion length in a light water media using MCNP code

Document Type : Scientific Note

Author

M. Arkani

Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.Box: 14155-1339, Tehran - Iran

https://doi.org/10.24200/nst.2022.1128.1745
Abstract
One of the key parameters of neutron diffusing media is the thermal neutron diffusion length. The popular calculational method is founded on utilizing MCNP code and curve fitting of a mathematical function to the neutron flux distribution in the media. In this investigation, a novel method based on the PTRAC card is proposed. The thermal neutron diffusion length parameter for light and water is calculated based on the above methods. The results are compared with the reported values and significant agreement is seen. The advantage of the above method is its independence from curve fitting to calculate the neutron flux distribution. In addition, in the curve fitting method, it is assumed that there is enough distance from the source location. This affects the error of the results because the neutron flux distribution function is correct everywhere except at the neutron source location. In the present investigation, the correctness of the proposed method in light water media is evaluated numerically.

Highlights

  1. John R. Lamarsh, Introduction to Nuclear Reactor Theory, Addison-Wesley Publishing Company, (1961).

 

  1. Denise B. Pelowitz, MCNPX User’s Manual, Version 2.6, (2008).

 

  1. E. Martinho, J. Salgado, Diffusion and extrapolation lengths of thermal neutrons in water by a stationary method, Journal of Nuclear Energy, 22(10), 597-600. (1968).

 

  1. V. Bhushan, S.K. Trikha, Pulsed neutron and diffusion length study in bismuth, Journal of Nuclear Energy, 25(1), 41-50 (1971).

 

  1. S.A.M.M. Siddiqui, M.M.R. Williams, A study of neutron diffusion lengths in heterogeneous, non-multiplying plate assemblies, Journal of Nuclear Energy, 27(5), 273-301 (1973).

 

  1. J. Csikai, A. Daroczy, K. Dede, Measurements of the diffusion length of thermal neutrons in water from 16 to 89° C and in diphyl (Dowtherm A) at 185° C, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 15(4), 204-208 (1961).

 

  1. M. Reier, The diffusion length of thermal neutrons in poisoned water, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 14(1-4), 186-188 (1961).

 

  1. I.C. Goyal, A.K. Ghatak, Diffusion length calculations in water, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 20(8), 659-665 (1966).

 

9. MATLAB 2020b, The MathWorks, Inc., Natick, Massachusetts, United State.

Keywords

MCNP code
Neutron diffusion length
Neutron flux
  1. John R. Lamarsh, Introduction to Nuclear Reactor Theory, Addison-Wesley Publishing Company, (1961).

 

  1. Denise B. Pelowitz, MCNPX User’s Manual, Version 2.6, (2008).

 

  1. E. Martinho, J. Salgado, Diffusion and extrapolation lengths of thermal neutrons in water by a stationary method, Journal of Nuclear Energy, 22(10), 597-600. (1968).

 

  1. V. Bhushan, S.K. Trikha, Pulsed neutron and diffusion length study in bismuth, Journal of Nuclear Energy, 25(1), 41-50 (1971).

 

  1. S.A.M.M. Siddiqui, M.M.R. Williams, A study of neutron diffusion lengths in heterogeneous, non-multiplying plate assemblies, Journal of Nuclear Energy, 27(5), 273-301 (1973).

 

  1. J. Csikai, A. Daroczy, K. Dede, Measurements of the diffusion length of thermal neutrons in water from 16 to 89° C and in diphyl (Dowtherm A) at 185° C, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 15(4), 204-208 (1961).

 

  1. M. Reier, The diffusion length of thermal neutrons in poisoned water, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 14(1-4), 186-188 (1961).

 

  1. I.C. Goyal, A.K. Ghatak, Diffusion length calculations in water, Journal of Nuclear Energy, Parts A/B. Reactor Science and Technology, 20(8), 659-665 (1966).

 

9. MATLAB 2020b, The MathWorks, Inc., Natick, Massachusetts, United State.

Journal of Nuclear Science, Engineering and Technology (JONSAT)
Volume 44, Issue 3 - Serial Number 105
October 2023
Pages 165-170

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