Implantable californium-252 needles have been developed as neutron sources for radiotherapy (1). These needles, designed as a versatile replacement for radium needles, have several theoretical advantages. The therapeutic effectiveness of fast neutrons, with their high linear energy transfer, may be greater than that of gamma rays on an equal dose basis (2). Fission spectrum neutrons provide a more localized dose to the tissue volume under treatment because of their low penetrability. Consequently, larger doses to the tumor can be given without an attendant increase in dose to surrounding healthy tissue. Californium, which emits 2.33 × 106 neutrons/(sec.) (πg) by spontaneous fission and has an effective half-life of 2.566 years, may be prepared in a variety of source geometries. Sufficient 252Cf for clinical use could become available by 1975 at a cost of about one dollar per microgram (3). The potential hazard associated with container rupture and release of contaminating materials is insignificant compared to that of radium sources. Before californium can be clinically evaluated, detailed descriptions of the dose distribution near californium sources must be determined. Preliminary dosimetry studies have been reported (1). More comprehensive studies have now been completed and are presented in this paper as neutron and gamma-ray depth-dose curves and isodose charts for a typical 252Cf needle. Materials and Methods Neutron and gamma-ray doses were measured in close proximity to a 4.5-πg 262Cf needle having a diameter of 1.65 mm and an active length of 20 mm (Fig. 1). All dose rates were measured in a torso-shaped phantom (20 × 30 × 60 cm high) filled with tissue-equivalent (T-E) solution (4). Two systems, activation threshold materials and semiconductor diode dosimeters, measured the incident fast neutron fluence in the presence of thermal neutrons and gamma rays (5). An equivalent fast neutron dose was calculated, using the kerma/ fluence relationship described by Williamson and Mitacek (6). Activation threshold materials for detecting fast neutron fluence have been used for years in reactor technology and in dosimetry. An adaptation of the dosimetry technics of Wright, Hoy, and Splichal (7) was employed for these measurements, which simulate an in vivo situation. The threshold materials (foils), described in Table I, were selected for their response to neutrons in a desired energy range. If an effective cross section weighted for the incident neutron energy distribution is determined, these three materials can measure the incident neutron fluence between 2.0 eV and 8.0 MeV. The fluence calculated for this interval is converted to neutron dose by where φCu, φIn, and φS are the fluenees in their respective energy ranges, and the coefficients are the average kerma/fluence factor, weighted for the californium fission spectrum. The foil system described has a minimum detectability limit of about 10 rads.
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