Verification and validation of the MCNPX-PoliMi code for simulations of neutron multiplicity counting systems

S. D. Clarke, E. C. Miller, M. Flaska, S. A. Pozzi, R. B. Oberer, L. G. Chiang

Research output: Contribution to journalArticlepeer-review

17 Scopus citations


Neutron coincidence counting is widely used in nuclear safeguards. Simulations of these systems can be performed using Monte Carlo codes such as MCNPX to aid in calibration or measurement design. However, the MCNPX coincidence-counting routines treat particle histories individually, therefore the dead time of the acquisition electronics is not treated. The MCNPX-PoliMi code provides the ability to model detailed effects such as data-acquisition electronics and system dead times. A specialized post-processing code has been developed to interpret the collision-log file and determine the response of a 3He multiplicity counter. The MCNPX-PoliMi simulation provides the full neutron multiplicity distribution measured by the 3He tubes. This distribution is used to compute the singles, doubles, and triples rates which are the quantities used to determine 235U mass. MCNPX-PoliMi has previously been validated with passive multiplicity measurements. In this study, a detailed analysis of the measurement system operating in active mode is presented for uranium-oxide standards ranging from 0.5 to 4.0 kg with a Canberra JCC-51 active well coincidence counter. MCNPX-PoliMi calculations are also compared with MCNPX. The two codes agree to within 1% for the cases with negligible dead times. The simulations are validated with measurements performed at the Y-12 National Security Complex.

Original languageEnglish (US)
Pages (from-to)135-139
Number of pages5
JournalNuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment
StatePublished - Jul 21 2013

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Instrumentation


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