Dose planning with comparison to in vivo dosimetry for epithermal neutron irradiation of the dog brain

Relative biological effectiveness for epithermal neutron beam contaminated with fast neutrons in the linear accelerator-based boron neutron capture therapy system coupled to a solid-state lithium target

This study aimed to quantify the relative biological effectiveness (RBE) for epithermal neutron beam contaminated with fast neutrons in the accelerator-based boron neutron capture therapy (BNCT) system coupled to a solid-state lithium target. The experiments were performed in National Cancer Center Hospital (NCCH), Tokyo, Japan. Neutron irradiation with the system provided by Cancer Intelligence Care Systems (CICS), Inc. was performed. X-ray irradiation, which was assigned as the reference group, was also performed using a medical linear accelerator (LINAC) equipped in NCCH. The four cell lines (SAS, SCCVII, U87-MG and NB1RGB) were utilized to quantify RBE value for the neutron beam. Before both of those irradiations, all cells were collected and dispensed into vials. The doses of 10% cell surviving fraction (SF) (D10) were calculated by LQ model fitting. All cell experiments were conducted in triplicate at least. Because the system provides not only neutrons, but gamma-rays, the contribution from the gamma-rays to the survival fraction were subtracted in this study. D10 value of SAS, SCCVII, U87-MG and NB1RGB for the neutron beam was 4.26, 4.08, 5.81 and 2.72 Gy, respectively, while that acquired by the X-ray irradiation was 6.34, 7.21, 7.12 and 5.49 Gy, respectively. Comparison of both of the D10 values, RBE value of SAS, SCCVII, U87-MG and NB1RGB for the neutron beam was calculated as 1.7, 2.2, 1.3 and 2.5, respectively, and the average RBE value was 1.9. This study investigated RBE of the epithermal neutron beam contaminated with fast neutrons in the accelerator-based BNCT system coupled to a solid-state lithium target.

The TL analysis methods used to determine absorbed gamma doses in vivo for the BNCT patients treated at FiR 1

The gamma dose determination using thermoluminescent (TL) dosimeters in mixed neutron-gamma fields, such as in boron neutron capture therapy (BNCT), is difficult due to the thermal neutron sensitivity of the detectors; especially when equipment capable of glow curve analysis is not available. The two TL analysis methods used previously in Finnish BNCT to correct the measured TL signal to obtain absorbed gamma dose in vivo were studied and compared, and an enhanced method was introduced. The three TL methods were found surprisingly consistent despite, e.g. the rough estimate made in the first method.