Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz

Effect of zirconium oxide nanoparticles on thermal, optical, and radiation shielding properties of Bi2O3-B2O3-MnO2 glasses

This study has explored the DSC, UV-Vis absorption spectroscopy, gamma ray and neutron shielding properties of Bi2O3-B2O3-MnO2: ZrO2 glasses. It demonstrates a unique approach to photon shielding analysis using JENDL/PD-2016 photonuclear data and employs a validated spherical neutron model for neutron shielding. Five transparent glasses were prepared with the chemical composition (in mol%) of 29Bi2O3-70B2O3-(1-x)MnO2: xZrO2, and labeled as MZ0.00 (for x = 0), MZ0.25 (for x = 0.25), MZ0.50 (for x = 0.5), MZ0.75 (for x = 0.75) and MZ1.00 (for x = 1). The glass ceramic nature of the samples has been characterized by DTA thermograms. The glass forming ability parameters (Kgl, S & H) were found to be highest for the sample MZ1.00. The UV-Visible optical absorption spectra have been interpreted, and hence the cut-off wavelength (λcut-off) and optical band gap (Eo) were evaluated. The absorption spectra have revealed the co-existence of manganese ions in three stable valence states Mn4+, Mn3+ and Mn2+ in the samples. When ZrO2 nanoparticles were added in the composition up to x = 0.50 mol%, the red shift in the cut-off wavelength (λcut-off) with gradual shrinkage in optical band gap (Eo) has been observed. Also, the linear and non-linear optical parameters viz., refractive index (no), non-linear refractive index (n2), linear optical susceptibility (χ(1)) and non-linear optical susceptibility (χ(3)) have been evaluated. These parameters showcased that B-O, Bi-O, Mn-O, Zr-O, etc. bonds could be strengthened by subsequent reduction of polarization of the trivalent ions (B3+ ions, Bi3+ ions and Mn3+ ions) in the glass system at higher concentrations of ZrO2. Photoatomic and photonuclear attenuation studies portrayed that the sample MZ0.50 has the lowest photon shielding capability. The fast neutron effective removal cross section (ΣR) was observed to be the highest for the sample MZ1.00. Thus, these glasses can be used to design the thermally stable transparent glasses, tunable optical elements, and radiation shielding materials.

Improvement of neutron sensitivity for lithium formate EPR dosemeters: a Monte Carlo analysis

This work presents the computational analysis of the sensitivity improvements that could be achieved in lithium formate monohydrate (LFM) electron paramagnetic resonance (EPR) dosemeters exposed to neutron beams. Monte Carlo (MC) simulations were performed on LFM pellets exposed to neutron beams with different energy spectra at various depths inside a water phantom. Various computations were carried out by considering different enrichments of 6Li inside the LFM matrix as well as addition of different amounts of gadolinium oxide inside the pellet blend. The energy released per unit mass was calculated with the aim of predicting the increase in dose achievable by the addition of sensitizers inside the pellets. As expected, a larger amount of 6Li induces an increase of energy released because of the charged secondary particles (i.e. 3H ions and α-particles) produced after neutron capture. For small depths in water phantom and low-energy neutron spectra the dose increase due to 6Li enrichment is high (more than three orders of magnitude with respect to the case of with 7Li). In case of epithermal neutron beams the energy released in 6Li-enriched LFM compound is smaller but larger than in the case of fast neutron beams. On the other hand, the computational analysis evidenced that gadolinium is less effective than 6Li in improving neutron sensitivity of the LFM pellets. Discussion based on the features of MC transport code is provided. This result suggests that 6Li enrichment of LFM dosemeters would be more effective for neutron sensitivity improvement and these EPR dosemeters could be tested for dosimetric applications in Neutron Capture Therapy.

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.

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.

Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications

This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented.

Boron neutron capture therapy (BNCT) in Finland: technological and physical prospects after 20 years of experiences

Boron Neutron Capture Therapy (BNCT) is a binary radiotherapy method developed to treat patients with certain malignant tumours. To date, over 300 treatments have been carried out at the Finnish BNCT facility in various on-going and past clinical trials. In this technical review, we discuss our research work in the field of medical physics to form the groundwork for the Finnish BNCT patient treatments, as well as the possibilities to further develop and optimize the method in the future. Accordingly, the following aspects are described: neutron sources, beam dosimetry, treatment planning, boron imaging and determination, and finally the possibilities to detect the efficacy and effects of BNCT on patients.

Fluence correction factors and stopping power ratios for clinical ion beams

BACKGROUND

In radiation therapy, the principal dosimetric quantity of interest is the absorbed dose to water. Therefore, a dose conversion to dose to water is required for dose deposited by ion beams in other media. This is in particular necessary for dose measurements in plastic phantoms for increased positioning accuracy, graphite calorimetry being developed as a primary standard for dose to water dosimetry, but also for the comparison of dose distributions from Monte Carlo simulations with those of pencil beam algorithms.

MATERIAL AND METHODS

In the conversion of absorbed dose to phantom material to absorbed dose to water the water-to-material stopping power ratios (STPR) and the fluence correction factors (FCF) for the full charged particle spectra are needed. We determined STPR as well as FCF for water to graphite, bone (compact), and PMMA as a function of water equivalent depth, z(w), with the Monte Carlo code SHIELD-HIT10A. Simulations considering all secondary ions were performed for primary protons as well as carbon, nitrogen and oxygen ions with a total range of 3 cm, 14.5 cm and 27 cm as well as for two spread-out Bragg-peaks (SOBP). STPR as a function of depth are also compared to a recently proposed analytical formula.

RESULTS

The STPR are of the order of 1.022, 1.070, and 1.112 for PMMA, bone, and graphite, respectively. STPR vary only little with depth except close to the total range of the ion and they can be accurately approximated with an analytical formula. The amplitude of the FCF depends on the non-elastic nuclear interactions and it is unity if these interactions are turned off in the simulation. Fluence corrections are of the order of a percent becoming more pronounced for larger depths resulting in dose difference of the order of 5% around 25 cm. The same order of magnitude is observed for SOBP.

CONCLUSIONS

We conclude that for ions with small total range (z(w-eq) ≤3 cm) dosimetry without applying FCF could in principle be performed in phantoms of materials other than water without a significant loss of accuracy. However, in clinical high-energy ion beams with penetration depths z(w-eq) ≥3 cm, where accurate positioning in water is not an issue, absorbed dose measurements should be directly performed in water or accurate values of FCF need to be established.

Dose determination using alanine detectors in a mixed neutron and gamma field for boron neutron capture therapy of liver malignancies

Boron Neutron Capture Therapy for liver malignancies is being investigated at the University of Mainz. One important aim is the set-up of a reliable dosimetry system. Alanine dosimeters have previously been applied for dosimetry of mixed radiation fields in antiproton therapy, and may be suitable for measurements in mixed neutron and gamma fields.

MATERIAL AND METHODS

Two experiments have been carried out in the thermal column of the TRIGA Mark II reactor at the University of Mainz. Alanine dosimeters have been irradiated in a phantom and in liver tissue.

RESULTS

For the interpretation and prediction of the dose for each pellet, beside the results of the measurements, calculations with the Monte Carlo code FLUKA are presented here. For the phantom, as well as for the liver tissue, the measured and calculated dose and flux values are in good agreement.

DISCUSSION

Alanine dosimeters, in combination with flux measurements and Monte Carlo calculations with FLUKA, suggest that it is possible to establish a system for monitoring the dose in a mixed neutron and gamma field for BNCT and other applications in radiotherapy.