Acceleration of Monte Carlo Based Treatment Planning - Criteria when Adjoint Calculations are Faster

Up until now, treatment planning in Boron Neutron Capture Therapy is only performed using Monte Carlo-based techniques. The conventional radiotherapy community has become more interested in such techniques, as they are impressed by the accuracy that can be achieved with Monte Carlo calculations. However, a disadvantage of the method is still the long times needed to obtain results with reliable statistics. Although computer power has become faster and cheaper over the years, it is still impossible to calculate a hundred or more different beam positions within a few days, which is the time a treatment planner in BNCT normally needs to produce an acceptable plan. With more calculated beam positions, a better treatment plan can be composed which can maximise the dose in the tumours whilst sparing the organs at risk. In normal (forward) Monte Carlo calculations, the particles start at the beam opening and travel into the tissue where they may or may not hit a target, e.g. tumour, organ at risk. In adjoint Monte Carlo calculations, the particles start at the target and travel out of the tissue to where the information is recorded. This information can be translated as if the particle started at the place of recording. With this method, the same information is gathered as with forward Monte Carlo but instantly all around the irradiated tissue. In a realistic head phantom with 10 organs at risk and 10 tumours, the adjoint techniques are 1.8 to 3.3 times faster than the forward MC calculations when 1020 different orientations of a gamma beam with a diameter larger than 5 cm are simulated. In the case of a neutron beam, the adjoint technique is faster by 6.6 to 20 times, than the forward MC. In general, in the case of small diameter beams, adjoint MC calculations are only preferable for a large number of beams and a small number of regions of interest. For larger beam sizes, the adjoint method is more favourable than the forward calculations when there are fewer beams and/or many regions of interest.JRC.F.4-Safety of future nuclear reactor

Comparison of Adjoint and Forward Monte Carlo Based Treatment Planning for a Realistic Head Phantom

BNCT treatment planning can be improved by having the adjoint technique available in the Monte Carlo transport code. In adjoint MC, the simulated particles travel backwards instead of ¿forward¿. By speeding up the calculations, more beam positions can be investigated and thus a better plan can be composed. In a realistic head phantom with ten disseminated lesions in the brain, the adjoint method is more favourable than the forward calculations whenever larger beam diameters are applied.JRC.F.4-Safety of future nuclear reactor

Monte Carlo based Treatment Planning Systems for Boron Neutron Capture Therapy in Petten

Boron Neutron Capture Therapy (BNCT) is a bimodal form of radiotherapy for the treatment of tumour lesions. Since the cancer cells in the treatment volume are targeted with 10B, a higher dose is given to these cancer cells due to the 10B(n,α)7Li reaction, in comparison with the surrounding healthy cells. In Petten (The Netherlands), at the High Flux Reactor (HFR), a specially tailored neutron beam has been designed and installed. Over 30 patients have been treated with BNCT in 2 clinical protocols: a phase I study for the treatment of glioblastoma multiforme (GM) [1] and a phase II study on the treatment of malignant melanoma (MM) [2]. Furthermore, activities concerning the extra-corporal treatment of metastasis in the liver (from colorectal cancer) are in progress [3, 4]. The irradiation beam at the HFR contains both neutrons and gammas that, together with the complex geometries of both patient and beam set-up, demands for very detailed treatment planning calculations. A well designed Treatment Planning System (TPS) should obey the following general scheme: (1) a pre-processing phase (CT and/or MRI scans to create the geometric solid model, cross-section files for neutrons and/or gammas); (2) calculations (3D radiation transport, estimation of neutron and gamma fluences, macroscopic and microscopic dose); (3) post-processing phase (displaying of the results, iso-doses and -fluences). Treatment planning in BNCT is performed making use of Monte Carlo codes incorporated in a framework, which includes also the pre- and post-processing phases. In particular, the GM protocol used BNCT_rtpe [5], while the MM protocol uses NCTPlan [6]. In addition, an ad hoc Positron Emission Tomography (PET) based treatment planning system (BDTPS) has been implemented in order to integrate the real macroscopic boron distribution obtained from PET scanning [8]. BDTPS is patented and uses MCNP as the calculation engine. The precision obtained by the Monte Carlo based TPSs exploited at Petten is considered sufficient for the scope of the project. One draw-back of the TPS in BNCT, compared with conventional TPSs, is the speed of calculation. In order to accelerate MCNP, a special ‘speed tally’ was implemented by which the results are obtained up to 10,000 times faster. Current studies enhance the application of the program Scan2MCNP [9] in order to translate the CT-images into a MCNP(X) geometry with a smaller voxel size and the application of mesh tallies.JRC.F.3-High Flux and Future Reactor

Design and Testing of a Rotating, Cooled Device for Extra-Corporal Treatment of Liver Cancer by BNCT in the Epithermal Neutron Beam at the HFR Petten

As part of the joint project on extra-corporal treatment of liver cancer by BNCT between JRC Petten and the University Hospital Essen, a facility has been designed and built to contain the liver during its irradiation treatment at the HFR Petten. The design consists of a rotating spheroid shaped PMMA holder, manufactured to open at the equator and closed by screwing together, surrounded by PMMA and graphite blocks. A validation exercise has been performed regarding both the nuclear conditions and the physical conditions. For the former, activation foil sets of Au, Cu and Mn, were irradiated at positions inside the liver holder filled with water, whilst a second measurement campaign has been performed using gel dosimetry. For the physical test, it is required to operate (rotate) the facility for up to 4 hours and to maintain the liver at approximately 4oC. The latter test was performed using “cold gun sprays” that inject cold air near the liver holder. Both the nuclear and physical validation tests were performed successfully.JRC.F.3-High Flux and Future Reactor

Use of Gel Dosimetry to Characterise the Dose Distribution in the Spheroidal Holder for Liver Treatment at the HFR Petten

The preparation of the treatment of the human liver tumour starts from the evaluation of the dosimetry of the problem. Simulations with a simplified spheroidal model of the human liver and the definition of the neutron beam normally used for BNCT clinical trials in Petten have been performed. The choice and the main properties of this liver facility is explained in another contribution to this conference. In order to validate the design calculations, which were performed using the Monte Carlo code MCNP, both activation foils and gel dosimeters have been used. The gel dosimeter is a technique to obtain continuous images of absorbed dose. By properly designing the gel isotopic composition, it is possible to separate the gamma dose and the dose due to charged particles, such as those produced in 10B reactions, and consequently the thermal neutron flux can be deduced. Therefore, this method gives an indication of the thermal neutron flux and the doses along pre-defined axes in the plane of the gel dosimeters, which have been positioned in the liver holder and surrounded with water. The calculated thermal neutron flux appears to be consistent with that obtained in strategic places of the liver holder, making use of the activation foils. The gamma and boron dose mappings obtained by the gel dosimeters confirm that the BNCT liver facility in Petten is able to provide a proper homogeneous thermal neutron distribution in the liver, as required for the successful treatment of the liver metastases.JRC.F.3-High Flux and Future Reactor

Monte Carlo Simulations of the Current Obtained with Ionisation Chamber Detectors in Mixed Fields of Neutrons and Gammas

It needs no introduction that good measurements regarding BNCT dosimetry are of vital interest. Above all, calculations for patient treatment planning are initially based on these measurements. Closely related, well understood dosimetry of mixed neutron and gamma fields is necessary to explain the outcomes of the many experiments performed. It is believed that the sometimes confusing and incomprehensible outcomes in BNCT research are due to incorrect dosimetry, i.e. misleading measurements. A popular detector used to describe the absorbed neutron and gamma doses is the ionisation chamber. To understand better the behaviour and intricacies of this detector, the collected and measured current is directly simulated with MCNPX. This Monte Carlo code is able to track neutrons, gammas and electrons all around and in the ionisation chamber. The calculated dose deposited by the electrons in the gas is proportional to the current measured. Protons and alphas emanating from the wall and/or gas materials due to nuclear reactions can also cause ionisations and thus add to the current. A custom-made program has been written to simulate this contribution. The issue in this study is that a disagreement between simulated and measured current can be caused by the computer code and/or measurement set-up and/or unknown influences of source and/or materials. Therefore, the model of the ionisation chamber as well as the neutron and gamma source descriptions are validated step-by-step. After having obtained enough confidence in the model it can be concluded that ionisation chamber measurements can be significantly affected by neutron interactions (this is energy dependent). Neutrons can increase the measured current due to unknown and unconsidered beta-, proton- and/or alpha-producers in the wall material and gas; this dose component does not exist without the presence of the ionisation chamber.JRC.F.4-Safety of future nuclear reactor

Validating a MCNPX Model of Mg(Ar) and TE(TE) Ionisation Chambers Exposed to 60Co Gamma-rays

For an accurate determination of the absorbed doses in complex radiation fields (e.g. mixed neutron-gamma fields), a better interpretation of the response of ionisation chambers is required. This study investigates a model of the ionisation chambers using a different approach, analysing the collected charge per minute as a response of the detector instead of the dose. The MCNPX Monte Carlo code is used. In this paper, the model is validated using a well-known irradiation field only: a 60Co source. The detailed MCNPX models of a Mg(Ar) and TE(TE) ionisation chamber is investigated comparing the measured charge per minute obtained free-in-air and in a water phantom with the simulated results. The difference between the calculations and the measurements for the TE(TE) chamber is within +2% whereas for the Mg(Ar) chamber is around 17%. The systematic discrepancy in the case of Mg(Ar) chamber is expected to be caused by an overestimation of the sensitive volume.JRC.F.3-Energy securit

Toward Prompt Gamma Spectroscopy for Monitoring Boron Distributions During Extra Corporal Treatment of Liver Metastases by Boron Neutron Capture Therapy - A Monte Carlo Simulation Study

A Monte Carlo calculation was carried out for Boron Neutron Capture Therapy (BNCT) of extra corporal liver phantom. The present paper describes the basis for a subsequent clinical application of the prompt gamma spectroscopy set up aimed at in vivo monitoring of boron distribution. MCNP code was used first to validate the homogeneity in thermal neutron field in the liver phantom and simulate the gamma rays detection system (collimator and detector) in the treatment room. The Gamma ray of 478 keV emitted by boron in small specific region can be detected and a mathematical formalism was used for the tomography image reconstruction.JRC.F.4-Safety of future nuclear reactor

Extension of the Calibration Curve for the PGRA Facility in Petten

At the boron neutron capture therapy (BNCT) facility in Petten, the Netherlands, 10B concentrations in biological materials are measured with the prompt gamma ray analyses facility that is calibrated using certified 10B solutions ranging from 0 to 210 ppm. For this study, newly certified 10B solutions ranging up to 1972ppm are added. MCNP simulations of the setup range to 5000 ppm. A second order polynomial (as already used) will fit 10B-concentrations less than 300 ppm. Above 300ppm a fitted third order polynomial is needed to describe the calibration curve accurately.JRC.F.4-Nuclear Reactor Integrity Assessment and Knowledge Managemen

Simulation of MG-AR and TE-TE Ionisation Chambers with MCNPX in a Straightforward Gamma and Beta Irradiation Field

New methods are required for a better interpretation of the response of ionisation chambers in order to improve further the determination of the absorbed doses in a mixed neutron-gamma BNCT field. Simulations might help to understand in particular the behaviour of the sensitivity factors of an ionisation chamber in the mixed field. The study presented here is a continuation of previous work with the final aim to obtain validated computer models of the Mg-Ar and TE-TE ionisation chambers. With these two chambers the so-called paired ionisation chamber technique is performed by which the neutron and gamma dose components can be separated and determined. By knowing exactly the neutron and gamma source in a simulation, the chamber response can be investigated. However, the validation of a simulated ionisation chamber set-up, starting directly with the mixed neutron/gamma fields is too complicated regarding the number of possibilities that could cause a discrepancy between measurements and simulations. Therefore, two simplified and well-known irradiation fields are considered first and will be discussed in this paper: an existing 60Co calibration source and a 90Sr check source which has been designed and constructed to perform ionisation chamber stability measurements. Both irradiation sources as well as the two ionisation chambers are modelled with MCNPX. This code has been used because it is capable of simulating neutrons among many other types of particles and rays. The model of the two ionisation chambers is investigated comparing the measured charge collection rate when the detectors are exposed in the 60Co gamma-ray field and in the 90Sr beta field with the calculated results. For the 60Co experiments the calculations agree within 3% with the measured values. For the 90Sr source the simulated Mg-Ar charge is 8% higher than the measurement and the simulated TE-TE charge is 6% lower than the measured charge from the ionisation chamber.JRC.F.4-Safety of future nuclear reactor