MCNP modelling of a combined neutron/gamma counter
A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of detector is to be used for high accuracy measurements of the plutonium content of nuclear materials. The multi-purpose Monte Carlo N-Particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO(2) and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study involving both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, to the (alpha,n), to spontaneous fission neutron emission rate ratio, and to the real coincidence gate utilization factor, are presented. Once bias-corrected, the trends of the real coincidence counts rate as a function of sample mass for three types of sample could be matched within 0.33% of experimental results. This confirms the possible use of MCNP to calculate response trends accurately for a wide variety of source materials, given a limited experimental calibration set.
Bibliographic Reference: Article: Nuclear Instruments and Methods in Physics Research Part A (1998)
Record Number: 199811180 / Last updated on: 1998-10-09
Original language: en
Available languages: en