Monte Carlo Modelling of Neutron Coincidence Counting Systems for Nuclear Safeguards

This paper describes Monte Carlo methodologies applied to the simulation of neutron coincidence or multiplicity counting systems, which are generally installed at a variety of nuclear facilities around the world and rather heavily relied on by Euratom and IAEA inspectors for the safeguards and non-proliferation of nuclear material (e.g. Mox fuel, PuO2 etc.) using Non Destructive Analysis (NDA) techniques. MCNP-PTA code has been developed at the Joint Research Centre of the European Commission in Ispra, in order to simulate the electronics pulse train analysis (PTA) systems that prevail in these counters in addition to the neutron radiation transport simulated by the MCNP part of the code on which it is primarily based. The methodology has been successfully applied for the design and optimization of new systems, for the numerical calibration and cross-calibration of various neutron counters and is to be deployed for the ¿verification¿ of nuclear materials by inspectors. This will considerably reduce both the measurement effort and the heavy reliance on reference materials required (but not always available) for the calibration of instruments. Agreement between Monte Carlo based calculations and measurements carried out either on site at nuclear installations or in laboratory conditions using reference materials are generally better than 3% as shown in this paper.

TAGZIRIA Hamid;  PEERANI Paolo; 

2010-10-19

American Nuclear Society (ANS)

JRC56443

https://publications.jrc.ec.europa.eu/repository/handle/JRC56443,