An environmental radionuclide baseline study near three Canadian naval ports.

This paper summarizes an environmental radionuclide baseline study undertaken for the Department of National Defence in Canada. The purpose of the project was to establish levels of radionuclides present in the environment around areas where nuclear propelled vessels may be berthed. Specifically, this report describes environmental baselines near Halifax (Nova Scotia), Esquimalt (British Columbia), and Nanoose Bay (British Columbia). Valued ecosystem component samples were taken from dairy farms, beef producers, market gardens, vegetables, tree fruits and berries within the study areas, as well as marine bivalves (mussels and clams), salmon, seaweed, and food from native fisheries. Numerous naturally occurring isotopes were detected and quantified. The only non-naturally occurring isotope positively identified was in the form of trace quantities of 131I, measured in the Halifax study zone (attributed to local hospital cancer therapy). 137Cs is the only other anthropogenic radionuclide detected. Its origin may be the combination of fallout from the Chernobyl accident and fallout from atmospheric nuclear weapon tests. The results indicate that nuclear-powered vessels have not resulted in activity levels that would contribute a significant radiation exposure to the public, the biota, and the environment within the three study zones.

Landmine detection: the problem and the challenge.

This paper explores the role of radiation methods in addressing the problem of detecting landmines. The application of neutron activation analysis, with an isotopic source or a pulsed neutron generator, is discussed. The use of neutron moderation as an indicator of the presence of a landmine is also explored. In addition, information provided by measuring scattered photons (gamma- and X-rays) is examined.

Experimental and computational determination of neutron dose equivalent around radiotherapy accelerators.

An experimental and computational investigation of the neutron dose equivalent in the treatment room and maze of various radiotherapy accelerators in Canada was completed. A bubble detector was selected for the bulk of the measurements due to its relative insensitivity to gamma radiation. The rooms and accelerator heads were modelled using the MCNP4B Monte Carlo radiation transport code, modified with a photoneutron patch. The results of the investigation showed generally good agreement between the experiments. Monte Carlo and analytical approximations to the neutron dose equivalent, and suggest that if the analytical approximations are used carefully they may be substituted for more costly experimental or Monte Carlo determinations.

Effectiveness of customised neutron shielding in the maze of radiotherapy accelerators.

An investigation was performed to examine the neutron dose equivalent in a radiotherapy maze lined with a customised neutron shielding material. The accelerator investigated was a Varian Clinac 2100C/D using 18 MV photons, and the neutron shielding utilised at this centre was Premadex commercially available neutron shielding. Based on Monte Carlo simulations, properly installed customised neutron shielding may reduce the neutron dose equivalent by up to a factor of 8 outside the maze, depending upon the installation. In addition, it was determined that the neutron dose near the entrance to the maze may be reduced by approximately 40% by using customised neutron shielding in the maze, as compared with a facility not using this shielding. This would have a positive dose-saving effect in doorless maze designs.

Intercomparison of Monte Carlo radiation transport codes to model TEPC response in low-energy neutron and gamma-ray fields.

Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities.