Neutron dosimetry, moderated energy spectrum, and neutron capture therapy for [formula omitted] medical sources

Abstract

[formula omitted] dosimetry was examined by calculative and experimental means. The Monte Carlo N-Particle (MCNP) transport code was used in a distributed computing environment (PVM) to determine neutron dose and neutron energy spectrum from [formula omitted] in a variety of clinically relevant materials. A Maxwellian spectrum was used to model [formula omitted] neutron emissions in these materials. Mixed-field dosimetry of [formula omitted] applicator tube (AT) sources was measured using tissue-equivalent ion chambers and a miniature GM counter to formulate a dosimetry protocol. Neutron dose was determined using the AAPM TG-43 dosimetry protocol. Results demonstrate the overwhelming dependence of dosimetry on the source geometry and no significant neutron attenuation by the source or encapsulation. Gold foils and TLDs were used to measure thermal neutron flux near [formula omitted] AT sources and compared with MCNP results. The fast neutron energy spectrum did not change markedly at greater distances from the AT source. Calculations of moderated [formula omitted] neutron energy spectrum with various loadings of [formula omitted] and [formula omitted] were performed, in addition to analysis of neutron capture therapy dosimetry. Radiological concerns such as personnel exposure and shielding of [formula omitted] emissions were examined. Feasibility of a high specific-activity HDR source was investigated through radiochemical and metallurgical studies using [formula omitted] stand-ins such as Tb, Gd, and [formula omitted] Issues such as capsule burst strength due to helium production for a variety of proposed HDR sources were addressed. At least 1 mg of [formula omitted] was necessary for a [formula omitted] HDR source. © 1999, American Association of Physicists in Medicine. All rights reserved.