ِِِِMonte Carlo simulation of a neutron irradiating system for determining efficiency of thermal neutron detectors

Authors

10.22052/5.2.21

Abstract

In this work, a neutron irradiating system containing six Am/Be neutron sources in a graphite moderator, which produces a relatively high and uniform thermal neutron flux in an irradiating cavity at the center of the system , has been  simulated .The thermal neutron flux in the center of the system has then been calculated and compared with experimental results  measured in a cavity at the center of a similar system .Good agreement between calculated and measured values show that true value of thermal neutron flux in this system (or any other irradiating system) can be accurately calculated by simulation method and be used  for calibrating neutron detectors and dosimeters. This system has been simulated using the MCNP Monte Carlo computer code. .In the calculations, thermal neutron flux has been first calculated at various distances from a single neutron source in a graphite moderator and then calculations have been performed and results discussed  for various cases of 2 , 4 and 6 neutron sources being present in the system. In addition, to investigate the effect of moderator properties on neutron flux distribution in the system, parallel calculations have been done for watermoderator, which show water is not a proper moderator for such a system.
 

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References

[1] M.G. Shahani, D. Sharma. Establishment of a standard thermal-neutron flux density at the BHABHA Atomic Research Centre. (1969). [2] V. Lacoste,V. Gresseir, H. Muller and L. Lebreton. Characterisation of the IRSN graphite moderated Americium-Berylium neutron field. Radiation Protection Dosimetry 110 (2004) 135–139. [3] E. Gallego, A. Lorente and H.R. Vega-Carrillo. Characteristics of the neutron field of the facility at DIN-UPM. Radiation Protection Dosimetry 110 (2004) 73–79. [4] H. Schuhmacher. International Standards Organization. Neutron calibration facilities. Radiation Protection Dosimetry110 (2004) 33–42. [5] M. Yoshizawa, S. Shimizu, Y. Kajimoto, T. Kawasaki, K. Fujii, J. Saegusa, Y. Tanimura and H. Yamamoto. Present status of calibration facility of JAERI-facility of radiation standards. http://www.ipen.br/biblioteca/cd/irpa/2004/files/3b46.pdf. [6] V. Lacoste. Design of a new thermal neutron field facility using Mont Carlo simulations. Radiation Protection Dosimetry (2007) 1–6. [7] T. Filho, R. Lima. Experimental and Monte Carlo evaluation of the neutron flux of an assembly with two Am-Be sources. Radiation Protection Dosimetry115 (2005) 412–414. [8] J. Zevallos-Chavez, C. Zamboni. Evaluation of the neutron flux distribution in an Am-Be irradiator using The MCNP-4C code. Brazilian Journal of Physics 353B (2005). [9] MCNPX User’s Manual,Version 2.4.0. LA-CP-02-408 September (2002). [10]http://www.srstechnol.com/QSA%20Catalogue/2004_Sources_Catalog_Version_2004–1112. pdf. [13] R. Pupheli. Purifying Coal for The Production of Nuclear Graphite. University of Pretoria, (2007). [14] A. Jones, G. Hall. Microstructural characterisation of nuclear grade graphite. Journal of Nuclear Materials 381 (2008) 152–158.

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