Answer to Question #2476 Submitted to "Ask the Experts"

Q

I have a piece of lithium glass (GS-20) (3" in diameter and 3 mm in thickness) with which I can count neutrons. I want to find a calibration method for this equipment to convert the neutron counts into neutron dose. How should I proceed? I have the MCNP (Monte Carlo N-Particle Transport Code); can I simulate the conversion factor? Or, does it has to be done experimentally?

A

First, neutron dosimetry and measurements are no doubt worthy of a half a semester of a graduate-level health physics course. Nonetheless, the Li-6 neutron–alpha reaction is often used in enriched Li-6 loaded glass, or TLD materials for neutron detection. The interaction cross section will vary decreasingly by several orders of magnitude with increasing energy. Also, the neutron interaction, initial energy, with flux, will have a direct bearing dose. And in practice, one never has a monoenergetic neutron source to measure, so polyenergetic neutrons in some spectrum must be evaluated. Thus, one needs to design a moderator/detector assembly that will thermalize the higher-energy neutrons so they can be detected. But the detected flux at each energy interval will contribute differently to the absorbed dose and, when a radiation weighting factor is applied for biological effectiveness, also to the dose equivalent.

You should read through Knoll's text (noted as a reference in this topical area of ATE) on the subject first. Information on the Li glass cross section could be used to estimate response with the moderator you may wish to use. This could be simulated with the MCNP code, but you'd also want to benchmark that response with an experimental rem responding instrument measurement for the given radiation field you're interested in monitoring. See our Ask the Expert Q2477 also.

David J. Allard, CHP