Monte Carlo model of the Studsvik BNCT clinical beam: Description and validation

Using ANN for thermal neutron shield designing for BNCT treatment room

Occupational radiation protection should be applied to the design of treatment rooms for various radiation therapy techniques, including BNCT, where escaping particles from the beam port of the beam shaping assembly (BSA) may reach the walls or penetrate through the entrance door. The focus of the present study is to design an alternative shielding material, other than the conventional material of lead, that can be considered as the material used in the door and be able to effectively absorb the BSA neutrons which have slowed down to the thermal energy range of < 1 eV after passing through the walls and the maze of the room. To this aim, a thermal neutron shield, composed of polymer composite and polyethylene, has been simulated using the Geant4 Monte Carlo code. The neutron flux and dose values were predicted using an artificial neural network (ANN), eliminating the need for time-consuming Monte Carlo simulations in all possible suggestions. Additionally, this technique enables simultaneous optimization of the parameters involved, which is more effective than the traditional sequential and separate optimization process. The results indicated that the optimized shielding material, chosen through ANN calculations that determined the appropriate thickness and weight percent of its compositions, can decrease the dose behind the door to lower than the allowable limit for occupational exposure. The stability of ANN was tested by considering uncertainties with the Gaussian distributions of random numbers to the testing data. The results are promising as they indicate that ANNs could be used as a reliable tool for accurately predicting the dosimetric results, providing a drastically powerful alternative approach to the time-consuming Monte Carlo simulations.

Optimized beam shaping assembly for a 2.1-MeV proton-accelerator-based neutron source for boron neutron capture therapy

Boron Neutron Capture Therapy (BNCT) is facing a new era where different projects based on accelerators instead of reactors are under development. The new facilities can be placed at hospitals and will increase the number of clinical trials. The therapeutic effect of BNCT can be improved if a optimized epithermal neutron spectrum is obtained, for which the beam shape assembly is a key ingredient. In this paper we propose an optimal beam shaping assembly suited for an affordable low energy accelerator. The beam obtained with the device proposed accomplishes all the IAEA recommendations for proton energies between 2.0 and 2.1 MeV. In addition, there is an overall improvement of the figures of merit with respect to BNCT facilities and previous proposals of new accelerator-based facilities.

Interaction between the biological effects of high- and low-LET radiation dose components in a mixed field exposure

PURPOSE

The relative biological effectiveness of two epithermal neutron sources, a reactor based source at Studsvik, Sweden, and a proton accelerator-based source in Birmingham, UK, was studied in relation to the proportional absorbed dose distribution as a function of neutron energy. Evidence for any interactions between the effects of biological damage induced by high- and low-linear energy transfer (LET) dose components, in this 'mixed field' irradiation, was also examined

MATERIALS AND METHODS

Clonogenic survival in Chinese Hamster-derived V79 cells was used to assess biological effectiveness in this study. Cells were irradiated in suspension at 4 °C at depths of 20, 35, 50 and 65 mm in a water phantom. This prevented the repair of sublethal damage, predominantly that produced by both incident and induced γ-rays in the field, over the variable periods of exposure required to irradiate cells with the same total absorbed dose. Cell survival, as a function of the absorbed radiation dose and depth in the phantom, was compared with Monte Carlo N-Particle (MCNP) calculations of the proportional absorbed dose distribution as a function of neutron energy for the two sources.

RESULTS

In terms of the dose-related reduction in clonogenic cell survival, the epithermal neutron source at Studsvik was more biologically effective than the Birmingham source at all depths considered in the phantom. Although the contribution from the high-LET dose component was greater for the Studsvik source at 20 mm depth in the phantom, at greater depths the dose contribution from the high-LET dose component at Studsvik overlap with those for the Birmingham source. However, the most striking difference is in the fast neutron component to the dose of the two sources, neutron energies>1 MeV were only associated with the Studsvik source. The relative biological effectiveness (RBE) of both sources declined slightly with depth in the phantom, as the total high-LET dose component declined. The maximum source RBE for Studsvik was 2.70±0.50 at 20 mm; reduced to 2.10±0.35 at depths of 50 and 65 mm. The corresponding values for Birmingham were 1.68±0.25 and 1.31±0.19, all values relate only to the surviving fraction of V79 cells at 37%, since RBE values are only applicable to the selected endpoint. Based on a dose reduction factor (DRF) of 1.0 for the total low-LET component to the absorbed dose, the RBE values for the high-LET dose component (fast neutrons and induced protons from the nitrogen capture reaction) was 14.5 and 7.05 for the Studsvik and Birmingham neutron sources, respectively. This is well outside the range of RBE historically reported values for V79 cells for the same level of cell survival for fast neutrons. The calculation of RBE values, based on the proportional absorbed dose distribution as a function of neutron energy, from historical data, and using a RBE of 1.8 for the dose from the nitrogen capture reaction, suggests RBE values for the total high-LET dose component of 3.1-2.8 and 2.5-2.0 for Studsvik and Birmingham, respectively, values again declining with depth in the phantom.

CONCLUSIONS

The overall biological effectiveness of the mixed field irradiation from an epithermal neutron sources depends on the composition and quality of the different dose components. The experimentally derived RBE values for the total high-LET dose components in these 'mixed field' irradiations are well in excess of historical data for fast neutrons. The difference between the historically expected and the observed RBE values is attributed to the interactions between the damage produced by high- and low-LET radiation.

An investigation of boron neutron capture therapy for recurrent glioblastoma multiforme

Objectives - To explore the use of boron neutron capture therapy (BNCT) for patients with glioblastoma multiforme (GBM), recurring after surgery and conventional radiotherapy (photon radiotherapy). Materials and Methods - Boron uptake in recurrent GBM was measured for four patients. Twelve patients were subsequently treated by BNCT with boronophenylalanine-fructose (900 mg/kg body weight), administered by intravenous infusion for 6 h. Results - Median survival time from initial diagnosis was 22.2 months. Comparison with other BNCT studies indicates a clinical advantage of the prolonged infusion. BNCT was well tolerated and quality of life remained stable until tumor progression for all 12 patients. No correlation was found between survival times and minimum tumor dose and number of radiation fields. Conclusions - Boron neutron capture therapy, with the prolonged procedure for infusion, is at least as effective as other radiation therapies for recurrent GBM and is delivered in one treatment session, with low radiation dose to the healthy brain. Survival from diagnosis compares favorably with that obtained with conventional radiotherapy plus concomitant and adjuvant temozolomide (TMZ) and survival from recurrence compares favorably with that obtained with TMZ at first relapse. The results of the present investigation are encouraging and should be confirmed in a randomized trial.

Monte Carlo calculations of thermal neutron capture in gadolinium: A comparison of GEANT4 and MCNP with measurements

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

Gadolinium neutron capture brachytherapy (GdNCB), a new treatment method for intravascular brachytherapy

Restenosis is a major problem after balloon angioplasty and stent implantation. The aim of this study is to introduce gadolinium neutron capture brachytherapy (GdNCB) as a suitable modality for treatment of stenosis. The utility of GdNCB in intravascular brachytherapy (IVBT) of stent stenosis is investigated by using the GEANT4 and MCNP4B Monte Carlo radiation transport codes. To study capture rate, Kerma, absorbed dose and absorbed dose rate around a Gd-containing stent activated with neutrons, a 30 mm long, 5 mm diameter gadolinium foil is chosen. The input data is a neutron spectrum used for clinical neutron capture therapy in Studsvik, Sweden. Thermal neutron capture in gadolinium yields a spectrum of high-energy gamma photons, which due to the build-up effect gives an almost flat dose delivery pattern to the first 4 mm around the stent. The absorbed dose rate is 1.33 Gy/min, 0.25 mm from the stent surface while the dose to normal tissue is in order of 0.22 Gy/min, i.e., a factor of 6 lower. To spare normal tissue further fractionation of the dose is also possible. The capture rate is relatively high at both ends of the foil. The dose distribution from gamma and charge particle radiation at the edges and inside the stent contributes to a nonuniform dose distribution. This will lead to higher doses to the surrounding tissue and may prevent stent edge and in-stent restenosis. The position of the stent can be verified and corrected by the treatment plan prior to activation. Activation of the stent by an external neutron field can be performed days after catherization when the target cells start to proliferate and can be expected to be more radiation sensitive. Another advantage of the nonradioactive gadolinium stent is the possibility to avoid radiation hazard to personnel.