Physical dosimetry and spectral characterization of neutron sources for neutron capture therapy

Abstract

Neutron beams useful for BNCT applications must be carefully characterized with regard to their free-field spectrum and intensity as well as with regard to the various dose components that are induced in tissue volumes under irradiation. The macroscopic physical characterization of a given beam is accomplished computationally by way of sophisticated neutron and photon transport analysis, validated by suitable experimental measurements. Such measurements are carried out by a variety of techniques. Neutron activation spectrometry may be viewed as the most accurate and reproducible approach for the measurement of the free-field neutron spectrum as well as for flux measurements in phantoms exposed to the beam of interest. This technique is based on the fact that different elements and different isotopes of the same element generally have different, more or less linearly independent, neutron activation responses as functions of incident neutron energy. In the case of free-field spectral measurements, a set of activation responses for 8-12 different materials having well-known activation cross sections as functions of energy is typically used in conjunction with a process for unfolding a neutron spectrum that is a best estimate (usually in a least-squares sense) of the true neutron spectrum that produced the observed set of activation responses. Activation techniques can also be used for spectral measurements of the neutron beam as it is modified by passage through phantom materials or tissue. The results can then be converted to absorbed physical dose via application of appropriate conversion factors for each neutron-induced dose component. Separate measurement of the incident and induced photon dose components that are inevitably present in the irradiation volume in BNCT may be accomplished by the use of paired ion-chamber technique, whereby one chamber is more sensitive to neutrons, while the other is more sensitive to photons. This also produces a background neutron-induced dose measurement that complements what can be obtained by activation spectrometry. Thermoluminescent dosimeters may also be used in a similar manner to separate the photon and neutron components of the background neutron dose. In addition to these basic approaches, several other types of radiation measurement instruments and techniques can be applied to provide additional dosimetric information. These include fission chambers, self-powered neutron detectors, and BF 3 and 3 He detectors for beam intensity monitoring, as well as proton-recoil chambers, solid organic crystal and plastic scintillators, and superheated nucleation detectors to provide additional spectral information. Finally, applications of Fricke dosimetry in tissue-equivalent gels have been explored as a possible means of two- and three-dimensional dosimetry in BNCT.