Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

Determining a methodology of dosimetric quality assurance for commercially available accelerator-based boron neutron capture therapy system

The irradiation field of boron neutron capture therapy (BNCT) consists of multiple dose components including thermal, epithermal and fast neutron, and gamma. The objective of this work was to establish a methodology of dosimetric quality assurance (QA), using the most standard and reliable measurement methods, and to determine tolerance level for each QA measurement for a commercially available accelerator-based BNCT system. In order to establish a system of dosimetric QA suitable for BNCT, the following steps were taken. First, standard measurement points based on tissue-administered doses in BNCT for brain tumors were defined, and clinical tolerances of dosimetric QA measurements were derived from the contribution to total tissue relative biological effectiveness factor-weighted dose for each dose component. Next, a QA program was proposed based on TG-142 and TG-198, and confirmed that it could be assessed whether constancy of each dose component was assured within the limits of tolerances or not by measurements of the proposed QA program. Finally, the validity of the BNCT QA program as an evaluation system was confirmed in a demonstration experiment for long-term measurement over 1 year. These results offer an easy, reliable QA method that is clinically applicable with dosimetric validity for the mixed irradiation field of accelerator-based BNCT.

Utilizing neutron generators in boron neutron capture therapy

Neutron capture therapy (NCT) is a radiotherapeutic technique that is designed to utilize the neutron capture reaction and damage the tumor cells through the energy release from the reaction. Nuclear reactors are typically utilized in this therapy because of the high neutron fluence rate that can be achieved. There has been minimal work to evaluate the effectiveness of neutron generators in NCT. This work presents the preliminary simulation results of utilizing of a deuterium-deuterium generator in boron neutron capture therapy. MCNP 6.1 was used to model the detailed geometry of the neutron generator and the phantom. Neutron moderators and photon shielding were used to optimize the neutron fluence rate in the tumor and decrease the photon dose in the phantom respectively. The study showed that a good localization of the neutron dose can be achieved in the tumor area with a reduction of the photon dose in the surrounding areas.

Adaption of a PIN-diode detector as an online neutron monitor for the thermal column of the TRIGA research reactor

A BNCT online neutron monitoring system was tested in a TRIGA reactor, using a silicon PIN-diode with a conversion foil. The setup was tested with different reactor powers at the hot and cold ends of the irradiation channel, using activation foils to compare with measured fluxes. The results demonstrate good reproducibility and show a linear correlation between signal of the PIN-diode and neutron flux at all positions, demonstrating this approach to be suitable for online monitoring of the neutron flux.

The alanine detector in BNCT dosimetry: dose response in thermal and epithermal neutron fields

PURPOSE

The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed.

METHODS

Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model.

RESULTS

The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields.

CONCLUSIONS

The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.

Comparison of EPR response of alanine and Gd₂O₃-alanine dosimeters exposed to TRIGA Mainz reactor

In this work we report some preliminary results regarding the analysis of electron paramagnetic resonance (EPR) response of alanine pellets and alanine pellets added with gadolinium used for dosimetry at the TRIGA research reactor in Mainz, Germany. Two set-ups were evaluated: irradiation inside PMMA phantom and irradiation inside boric acid phantom. We observed that the presence of Gd2O3 inside alanine pellets increases the EPR signal by a factor of 3.45 and 1.24 in case of PMMA and boric acid phantoms, respectively. We can conclude that in the case of neutron beam with a predominant thermal neutron component the addition of gadolinium oxide can significantly improve neutron sensitivity of alanine pellets. Monte Carlo (MC) simulations of both response of alanine and Gd-added alanine pellets with FLUKA code were performed and a good agreement was achieved for pure alanine dosimeters. For Gd2O3-alanine deviations between MC simulations and experimental data were observed and discussed.

Cellular uptake and in vitro antitumor efficacy of composite liposomes for neutron capture therapy

Background

Neutron capture therapy for glioblastoma has focused mainly on the use of ¹⁰B as neutron capture isotope. However, ¹⁵⁷Gd offers several advantages over boron, such as higher cross section for thermal neutrons and the possibility to perform magnetic resonance imaging during neutron irradiation, thereby combining therapy and diagnostics. We have developed different liposomal formulations of gadolinium-DTPA (Magnevist®) for application in neutron capture therapy of glioblastoma. The formulations were characterized physicochemically and tested in vitro in a glioma cell model for their effectiveness.

Methods

Liposomes entrapping gadolinium-DTPA as neutron capture agent were manufactured via lipid/film-extrusion method and characterized with regard to size, entrapment efficiency and in vitro release. For neutron irradiation, F98 and LN229 glioma cells were incubated with the newly developed liposomes and subsequently irradiated at the thermal column of the TRIGA reactor in Mainz. The dose rate derived from neutron irradiation with ¹⁵⁷Gd as neutron capturing agent was calculated via Monte Carlo simulations and set in relation to the respective cell survival.

Results

The liposomal Gd-DTPA reduced cell survival of F98 and LN229 cells significantly. Differences in liposomal composition of the formulations led to distinctly different outcome in cell survival. The amount of cellular Gd was not at all times proportional to cell survival, indicating that intracellular deposition of formulated Gd has a major influence on cell survival. The majority of the dose contribution arises from photon cross irradiation compared to a very small Gd-related dose.

Conclusions

Liposomal gadolinium formulations represent a promising approach for neutron capture therapy of glioblastoma cells. The liposome composition determines the uptake and the survival of cells following radiation, presumably due to different uptake pathways of liposomes and intracellular deposition of gadolinium-DTPA. Due to the small range of the Auger and conversion electrons produced in ¹⁵⁷Gd capture, the proximity of Gd-atoms to cellular DNA is a crucial factor for infliction of lethal damage. Furthermore, Gd-containing liposomes may be used as MRI contrast agents for diagnostic purposes and surveillance of tumor targeting, thus enabling a theranostic approach for tumor therapy.