Citation:
H. Kistle, “MC-15 Neutron Multiplicity Counter Characterization for Dosimetry and Spectroscopy Applications”, Ph.D. Dissertation, Nuclear Engineering, Texas A&M University, College Station, TX (2024).
Abstract:
This work evaluates the ability of an established and fielded neutron multiplicity detector, the MC-15, to produce accurate quantification of additional neutron field information beyond its existing neutron multiplicity analysis. The MC-15 is a field-deployable portable multiplicity counter.rnThe detector provides information regarding personnel risk and nuclear assessment. This work aims to supplement that information. Risk to personnel includes not only the criticality hazard posed by a multiplying medium (which is an existing analytical capability) but also the radiation hazard – the neutron dose rate. The MC-15 response was characterized relative to a calibrated survey meter to establish a conversion method from the detector count rate to neutron dose rate. The result yielded a conversion function dependent on the distance from the source to the MC-15 for fission-energy neutrons regardless of polyethylene shielding presence. In addition, the design characteristics of the MC-15 provide rough energy information relative to fission-energy neutrons, for which the MC-15 was designed. These characteristics can be further leveraged for improved knowledge of the source energy spectrum. Energy-dependent response functions were created from simulations of the MC-15 in two configurations. The response functions were used with an existing neutron energy spectrum deconvolution code to quantify the ability of the MC-15 to produce an adequate solution spectrum from both measured and simulated data. Although the standard measurement orientation is severely limited in its ability to provide spectroscopic analysis, orienting the detector sideways with respect to the item shows promise in producing supplemental energy information. Finally, further characterization of the detector efficiency response in its sideways orientation was used to generate an alternate method for neutron source strength (NSS) calculation. Simulation-based characterization of the MC-15 was conducted in a parametric study. The results of the characterization allow for correction of the bare efficiency calculation for the presence and amount of polyethylene moderation.