Evaluation of Neutron Detection Efficiency of the Unified Non-Destructive Assay Using MCNPX Code

MCNPX 코드를 이용한 통합비파괴측정장치의 중성자 검출 효율 평가

  • Received : 2013.10.15
  • Accepted : 2013.12.04
  • Published : 2013.12.30


In this study, neutron detection efficiency of the UNDA system, which has been developed for study on nuclear material accountancy in a future pyro-process facility, was evaluated by using the MCNPX code. The detection efficiency was evaluated as a function of (1) positions of $^{252}Cf$ neutron source in the axial and radial directions, and (2) thicknesses and locations of the container filled with the depleted uranium materials for two different designs of the UNDA. In the case of $^{252}Cf$ source positions, detection efficiency was distributed from 6.83% to 13.35%. As $^{252}Cf$ source was positioned at upper part in the axial direction, detection efficiency was decreased after a slight increase. On the other hands, as $^{252}Cf$ source was positioned at outer part in the radial direction, detection efficiency was increased. In the case of container thickness, there was a slight decline when the thickness was increased. As the container was located at upper part, detection efficiency was decreased and as the container was located at outer part, detection efficiency was increased. Detection efficiency was varied from 10.31% to 13.61%. These values were higher than that of $^{252}Cf$ source case. The UNDA with polyethylene cover has about 2% higher detection efficiency than the UNDA without the cover.


Supported by : 한국연구재단


  1. Pelowitz DB. MCNPX user's manual version 2.7.0. LA-CP-11-00438. Los Alamos National Laboratory. 2011.
  2. Verbeke JM, Hagmann C, Wright D. Simulation of neutron and gamma ray emission from fission and photofission. UCRL-AR-228518. Lawrence Livermore National Laboratory. 2010.

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