Impact of neutron sieve therapy for deep site in boron neutron capture therapy using Monte Carlo calculation.

Boron Neutron Capture Therapy (BNCT) is an advanced form of radiotherapy technique that is potentially superior to all conventional techniques for cancer treatment, as it is targeted at killing individual cancerous cells with minimal damage to surrounding healthy cells. After decades of development, BNCT has reached clinical-trial stages in several countries, mainly for treating challenging cancers such as malignant brain tumors. The Indonesian consortium of BNCT already developed of the design BNCT for many cases of type cancers using many neutron sources. The main objective of the Indonesian consortium BNCT are the development of BNCT technology package which consists of a non nuclear reactor neutron source based on cyclotron and compact neutron generator technique, advanced boron-carrying pharmaceutical, and user-friendly treatment platform with automatic operation and feedback system as well as commercialization of the BNCT though franchised network of BNCT clinics worldwide. The Indonesian consortium BNCT will offering to participate in Boron carrier pharmaceuticals development and testing, development of cyclotron and compact neutron generators and provision of neutrons from the 100 kW Kartini Research Reactor to guide and to validate compact neutron generator development. Studies were carried out to design a collimator which results in epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle 5 (MCNP5) codes. Reactor within 100 kW of output thermal power was used as the neutron source. The design criteria were based on the IAEA’s recommendation. All materials used were varied in size, according to the value of mean free path for each. Monte Carlo simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 1,5 cm thick of Bi as "-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 3-5 cm varied aperture size, epithermal neutron beam with minimum flux of 7,8 x 108 n.cm-2.s-1, maximum fast neutron and "-ray components of, respectively, 1,9 x 10-13 Gy.cm2.n-1 and 1,8 x 10-13 Gy.cm2.n-1, maximum thermal neutron per epithermal neutron ratio of 0,009, and beam minimum directionality of 0,72, could be produced. The beam did not fully pass the IAEA’s criteria, since the epithermal neutron flux was still below the recommended value, 1,0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeded 5 x 108 n.cm-2.s-1. When this collimator was surrounded by 8 cm thick of graphite, the characteristics of the beam became better that it passed all IAEA’s criteria with epithermal neutron flux up to 1,7 x 109 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment and study of many cases cancer type i.e.; liver and lung curcinoma. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Sodium boroncaptate (BSH) was used as in this research. BSH had effected in liver for radiation quality factor as 0.8 in health tissue and 2.5 in cancer tissue. Modelling organ and source used liver organ who contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 $g/g cancer. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Given the advantages of low density owned by lungs, hence BNCT is a solid option that can be utilized to eradicate the cell cancer in lungs. Modelling organ and neutron source for lung carcinoma was used Compact Neutron Generator (CNG) by deuterium-tritium which was used is boronophenylalanine (BPA). The concentration of boron-10 compound was varied in the study; i.e. the variations were 20; 25; 30; 35; 40 and 45 μg.g-1 cancer tissues. Ideally, the primary dose which is solemnly expected to contribute in the therapy is alpha dose, but the secondary dose; i.e. neutron scattering dose, proton dose and gamma dose that are caused due to the interaction of thermal neutron with the spectra of tissue can not be simply omitted. Thus, the desired output of MCNPX; i.e. tally, were thermal and epithermal neutron flux, neutron and photon dose. The liver study variation of boron concentration result dose rate to every variation were0,042; 0,050; 0,058; 0,067; 0,074; 0,082; 0,085 Gy/sec. Irradiation time who need to every concentration were 1194,687 sec (19 min 54 sec);999,645 sec (16 min 39 sec); 858,746 sec (14 min 19 sec); 743,810 sec (12 min 24 sec); 675,156 sec (11 min 15 sec); 608,480 sec (10 min 8 sec); 585,807sec (9 min 45 sec). The lung carcinoma study variations of boron-10 concentration in tissue resulted in the dose rate of each variables respectively were 0.003145, 0.003657, 0.00359, 0.00385, 0.00438 and 0.00476 Gy.sec-1 . The irradiated time needed for therapy for each variables respectively were 375.34, 357.55, 287.58, 284.95, 237.84 and 219.84 minutes.

In boron neutron capture therapy (BNCT)-a form of binary radiotherapy-the primary challenge in treatment planning systems for dose calculations arises from the time-consuming nature of the Monte Carlo (MC) method. Recent progress, including the use of neural networks (NN), has been made to accelerate BNCT dose calculations. However, this approach may result in significant dose errors in both the tumor and the skin, with the latter being a critical organ in BNCT. Furthermore, owing to the lack of physical processes in purely NN-based approaches, their reliability for clinical dose calculations in BNCT is questionable. In this study, a physically constrained MC-NN (PCMC-NN) coupling algorithm is proposed to achieve fast and accurate computation of the BNCT three-dimensional (3D) therapeutic dose distribution. This approach synergizes the high precision of the MC method with the speed of the NN and utilizes physical conservation laws to constrain the coupling process. It addresses the time-consuming issue of the traditional MC method while reducing dose errors. Clinical data were collected from 113 glioblastoma patients. For each patient, the 3D dose distributions for both the coarse and detailed dose grids were calculated using the MC code PHITS. Among these patients, the data from 14 patients were allocated to the test set, 9 to the validation set, and the remaining to the training set. A neural network, 3D-Unet, was built based on the coarse grid dose and patient CT information to enable fast and accurate computation of the 3D detailed grid dose distribution of BNCT. Statistical evaluations, including relative deviation, dose deviation, mean absolute error (MAE), and mean absolute percentage error (MAPE) were conducted. Our findings suggested that the PCMC-NN algorithm substantially outperformed the traditional NN and interpolation methods. Furthermore, the proposed algorithm significantly reduced errors, particularly in the skin and GTV, and improved computational accuracy (hereinafter referred to simply as 'accuracy') with a MAPE range of 1.6%-4.0% and a maximum MAE of 0.3Gy (IsoE) for different organs. The dose-volume histograms generated by the PCMC-NN aligned well with those obtained from the MC method, further validating its accuracy. The PCMC-NN algorithm enhanced the speed and accuracy of BNCT dose calculations by combining the MC method with the NN algorithm. This indicates the significant potential of the proposed algorithm for clinical applications in optimizing treatment planning.

Introduction: Boron neutron capture therapy (BNCT) is recommended to treat the glioblastoma tumor. It is well known that neuron beams are more effective treatment than photon beams to treat hypoxia tumors due to interaction of neutron with nucleus and production of heavy particles such as 7Li and alpha particle. In this study to evaluate the suitability of BNCT for treating of lung cancer, the dose distributions of neutron beam were calculated in lung tumor volume and in peripheral organs at risk (OARs). Materials and Methods: Dose distribution in Boron neutron capture therapy to treat lung cancer was calculated by MCNPX (2.6.0) code. A 3×3×3 cm3 tumor was located in left lung of ORNL phantom and was irradiated with a rectangular field of neutron positioned at surface source distance (SSD) of 10 cm. recommended spectrum of MIT (Massachusetts Institute of Technology) was used. Tumor was loaded with different concentrations of Boron 0, 10, 30 and 60 ppm. Dose delivered to OARs such as heart, spinal cord, right lung … were calculated. Results: The results show that neutron flux significantly decreased followed to penetrate in lung tissue. Neutron flux decreased in all energy bins of irradiated MIT spectrum; maximum fall- off occurred in the range of epithermal energy. Dose distribution was not depended to SSD. The absorbed dose in tumor was 2.16×10-14, 2.6×10-14, 3.44×10-14 and 4.72×10-14 Gy(per one irradiated neutron from source) for boron concentration of 0, 10, 30 and 60 ppm, respectively. From the OARs, the heart tissue absorbed the maximum dose of 1.66×10-15 Gy (per one irradiated neutron from source). Conclusion: Our simulated model was successful to calculated organ doses in BNCT. As the boron concentration in lung tumor increases, absorbed dose increased while dose uniformity trended downward. Our results show that the MIT neutron source is suitable to treat deep lung tumors while OVRs’ dose maintains within the threshold dose.

Boron Neutron Capture Therapy (BNCT) is a radiotherapy technique based on the enrichment of tumour cells with suitable 10-boron concentration and on subsequent neutron irradiation. Low-energy neutron irradiation produces a localized deposition of radiation dose caused by boron neutron capture reactions. Boron is vehiculated into tumour cells via proper borated formulations, able to accumulate in the malignancy more than in normal tissues. The neutron capture releases two high-LET charged particles (i.e., an alpha particle and a lithium ion), losing their energy in a distance comparable to the average dimension of one cell. Thus BNCT is selective at the cell level and characterized by high biological effectiveness. As the radiation field is due to the interaction of neutrons with the components of biological tissues and with boron, the dosimetry requires a formalism to express the absorbed dose into photon-equivalent units. This work analyzes a clinical case of an adenoid cystic carcinoma treated with carbon-ion radiotherapy (CIRT), located close to optic nerve and deep-seated as a practical example of how to apply the formalism of BNCT photon isoeffective dose and how to evaluate the BNCT dose distribution against CIRT. The example allows presenting different dosimetrical and radiobiological quantities and drawing conclusions on the potential of BNCT stemming on the clinical result of the CIRT. The patient received CIRT with a dose constraint on the optic nerve, affecting the peripheral part of the Planning Target Volume (PTV). After the treatment, the tumour recurred in this low-dose region. BNCT was simulated for the primary tumour, with the goal to calculate the dose distribution in isoeffective units and a Tumour Control Probability (TCP) to be compared with the one of the original treatment. BNCT was then evaluated for the recurrence in the underdosed region which was not optimally covered by charged particles due to the proximity of the optic nerve. Finally, a combined treatment consisting in BNCT and carbon ion therapy was considered to show the consistency and the potential of the model. For the primary tumour, the photon isoeffective dose distribution due to BNCT was evaluated and the resulted TCP was higher than that obtained for the CIRT. The formalism produced values that are consistent with those of carbon-ion. For the recurrence, BNCT dosimetry produces a similar TCP than that of primary tumour. A combined treatment was finally simulated, showing a TCP comparable to the BNCT-alone with overall dosimetric advantage in the most peripheral parts of the treatment volume. Isoeffective dose formalism is a robust tool to analyze BNCT dosimetry and to compare it with the photon-equivalent dose calculated for carbon-ion treatment. This study introduces for the first time the possibility to combine the dosimetry obtained by two different treatment modalities, showing the potential of exploiting the cellular targeting of BNCT combined with the precision of charged particles in delivering an homogeneous dose distribution in deep-seated tumours.

Introduction: The Monte Carlo simulation is used to enhance reliability in the experiments related to nuclear instruments. in addition, that is used to calculate the different components of the neutron and gamma ray fluxes in boron neutron capture therapy(BNCT) and neutron capture (NCT)applications. BNCT is one of the methods in radiotherapy, that is used the neutron beam for kill the cancer cells. The neutron activation analysis(NAA) is the method for identify light elements that the neutron is captured with light elements nucleus and emits characteristic gamma rays. Materials and Methods: the MCNPX code was used for calculation. Boron and other light elements exist in the liver tissue. The BNCT special set geometry was designed. In this designed, light elements analysis is performed simultaneously with the neutron therapy. The effective parameters such as source location, source type, detector location, detector material, patient couch, energy of source, moderator, collimator type, length and thickness of collimator, distance between sample and source, opening of collimator, geometry and location of detector was designed. Results: the best neutron source for BNCT and light element analysis is expanded neutron spectrum produced by the reactor with Paraffin moderator. Neutron Source Generator with every moderator had low efficiency. Collimator made of graphite, graphene and carbon compounds had better neutron output spectrum. Sodium iodide detector is suitable for the detection of light elements gamma rays. The collimator length 20 cm and thickness 6cm. The detectors are placed in a cylindrical arrangement and They should not be exposed to direct neutron radiation. Conclusion: the MCNP study is one of the best methods for BNCT and NCT. the NAA and BNCT is possible Performing Simultaneously. The expanded neutron spectrum from reactors is suitable for NAA and BNCT.

ID: 122 Verification of dose estimation for Monte-Carlo based treatment planning system for boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a binary treatment modality that uses high LET particles to achieve tumor cell killing. Deuterium-deuterium (DD) compact neutron generators have advantages over nuclear reactors and large accelerators as the BNCT neutron source, such as their compact size, low cost, and relatively easy installation. The purpose of this study is to design a beam shaping assembly (BSA) for a DD neutron generator and assess the potential of a DD-based BNCT system using Monte Carlo (MC) simulations. The MC model consisted of a head phantom, a DD neutron source, and a BSA. The head phantom had tally cylinders along the centerline for computing neutron and photon fluences and calculating the dose as a function of depth. The head phantom was placed at 4cm from the BSA. The neutron source was modeled to resemble the source of our current DD neutron generator. A BSA was designed to moderate and shape the 2.45-MeV DD neutrons to the epithermal (0.5eV to 10keV) range. The BSA had multiple components, including moderator, reflector, collimator, and filter. Various materials and configurations were tested for each component. Each BSA layout was assessed in terms of the in-air and in-phantom parameters. The maximum brain dose was limited to 12.5 Gray-Equivalent (Gy-Eq) and the skin dose to 18Gy-Eq. The optimized BSA configuration included 30cm of lead for reflector, 45cm of LiF, and 10cm of MgF2 for moderator, 10cm of lead for collimator, and 0.1mm of cadmium for thermal neutron filter. Epithermal flux at the beam aperture was 1.0×105 nepi /cm2 -s; thermal-to-epithermal neutron ratio was 0.05; fast neutron dose per epithermal was 5.5×10-13 Gy-cm2 /φepi , and photon dose per epithermal was 2.4×10-13 Gy-cm2 /φepi . The AD, AR, and the advantage depth dose rate were 12.1cm, 3.7, and 3.2×10-3 cGy-Eq/min, respectively. The maximum skin dose was 0.56Gy-Eq. The DD neutron yield that is needed to irradiate in reasonable time was 4.9×1013 n/s. Results demonstrated that a DD-based BNCT system could be designed to produce neutron beams that have acceptable in-air and in-phantom characteristics. The parameter values were comparable to those of existing BNCT facilities. Continuing efforts are ongoing to improve the DD neutron yield.

<bold>Background</bold>Boron Neutron Capture Therapy (BNCT) is an advanced and precise targeted tumor treatment technology with broad application prospects. One of the core tasks in the development of this technology is the measurement of neutron flux. The primary measurement methods for neutron energy spectrum and flux in existing BNCT radiation fields are the multi-foil activation method and the Bonner sphere spectrometer method. These methods have problems such as complicated operation procedures, long measurement time, and inability to measure in real time online. Therefore, new detection technologies need to be developed urgently.<bold>Purpose</bold>This study aims to apply Monte Carlo simulation to the design of a diamond-based epithermal neutron flux detector tailored for BNCT neutron beam measurement requirements to achieve real-time measurement of epithermal neutron flux.<bold>Methods</bold>The Monte Carlo simulation code MCNP5 was employed to design and optimize the detector structure, neutron response of the detector, selection of neutron conversion materials, selection of slowing materials, and high-energy neutron resonance absorption strategy. Equipped with a nuclear electronics readout system, the optimized super-thermal neutron detector was applied to real time measurement of superheated neutron flux, and performance verification of the BNCT radiation fields from accelerator neutron sources and reactor neutron sources were conducted.<bold>Results</bold>The optimized design for epithermal neutron flux detector is to use a D₂O sphere with a radius of 37 cm and a 0.01 cm Fe foil as the moderator and high-energy neutron absorber materials, the size of the diamond detector is 1.0 cm×1.0 cm×0.03 cm, and ⁶Li of 17 μm is selected as the neutron conversion layer. Verification results show that the neutron response energy range of the designed detector is 0.5 eV to 10 keV, and the uncertainty of the super-thermal neutron flux measurement results is less than 4%, which is better than the existing multi-film activation method and multi-ball spectrometer method.<bold>Conclusions</bold>Results of this study demonstrate that designed diamond epithermal neutron flux detector can accurately complete the super-thermal neutron flux measurement of the BNCT radiation field and has guiding significance for the subsequent fabrication and testing of the detector.

The Monte Carlo stochastic simulation technique has traditionally been the only well-recognized method for computing three-dimensional radiation dose distributions in connection with boron neutron capture therapy (BNCT) research. A deterministic approach to this problem would offer some advantages over the Monte Carlo method. This paper describes an application of a deterministic method to analytically simulate BNCT treatment of a canine head phantom using the epithermal neutron beam at the Brookhaven medical research reactor (BMRR). Calculations were performed with the TORT code from Oak Ridge National Laboratory (ORNL), an implementation of the discrete ordinates, or Sn method. Calculations were from first principles and used no empirical correction factors. The phantom surface was modeled by flat facets of approximately 1 cm2. The phantom interior was homogeneous. Energy-dependent neutron and photon scalar fluxes were calculated on a 32 x 16 x 22 mesh structure with 96 discrete directions in angular phase space. The calculation took 670 min on an Apollo DN10000 workstation. The results were subsequently integrated over energy to obtain full three-dimensional dose distributions. Isodose contours and depth-dose curves were plotted for several separate dose components of interest. Phantom measurements were made by measuring neutron activation (and therefore neutron flux) as a function of depth in copper-gold alloy wires that were inserted through catheters placed in holes drilled in the phantom. Measurements agreed with calculations to within about 15%. The calculations took about an order of magnitude longer than comparable Monte Carlo calculations but provided various conveniences, as well as a useful check.

Objective To design a scheme of epithermal neutron beam used for boron neutron capture therapy (BNCT).Methods Based on Tsinghua University experimental reactor and its No.1 passage,five schemes comprised of moderate materials,absorbing materials of thermal neutron and γ shielding materials were designed according to different locations of materials placed in No.1 passage.To select a proper scheme from five schemes,the neutron fluence rate,the neutron dose rate and γ dose rate at exit of beam in each scheme were calculated with Monte Carlo simulating methods and then contrasted with BNCT technique criterion.Results The scheme of epithermal neutron beam meeting technical requirements of BNCT was obtained,in which the thickness of moderate material,absorbing materials of thermal neutron and γ shielding materials are 53.5 cm,2 mm and 9 cm,respectively.Conclusions The theoretical scheme could provide some reference to realize BNCT on reactor. Key words: BNCT; Monte Carlo method; simulation and calculation; Epithermal neutron radiation field

Boron neutron capture therapy (BNCT) is expected to be an effective method of improving the treatment results on malignant brain glioma and malignant melanoma, for which no successful treatment has been developed so far. The beam shaping assembly (BSA) of accelerator-based boron neutron capture therapy (A-BNCT) consists of a moderator, a reflector, gamma and thermal neutron shielding and a collimator. The BSA moderates the fast neutron produced in target to epithermal energy range. Design of BSA is one of the key jobs in BNCT project. An optimized study was conducted to design a beam shaping assembly for BNCT facility based on 3.5 MeV 10 mA radio-frequency quadrupole proton accelerator at Dongguan Neutron Science Center. In this simulation work, the neutron produced from the 7Li (p, n) 7Be reaction by 3.5 MeV proton is adopted as a neutron source term. In order to search for an optimized beam shaping assembly for accelerator-based BNCT, Monte Carlo simulation is carried out based on the parameters of moderator material and structure, the Gamma shielding, and the thermal neutron filter in the beam shaping assembly. The beam shaping assembly in this work consists of various moderator materials, teflon as reflector, Bi as gamma shielding, 6Li as thermal neutron filter, and lithium polyethylene as collimator. After comparing the simulation results of Fluental and LiF moderator materials, the beam shaping assembly configuration based on sandwich Fluental-LiF configuration is proposed. The sandwich Fluental-LiF configuration is made up of Fluental and LiF layer by layer, like a sandwich structure, and each layer is 2 cm thick. According to the beam quality requirement of the IAEA-tecdoc-1223 report, the optimized epithermal neutron flux in air at the exit of BSA of the sandwich Fluental-LiF configuration is 9.14×108 n/(cm2·s), which is greater than those of the Fluental configuration (7.81×108 n/(cm2·s)) and LiF configuration (8.79×108 n/(cm2·s)), when the ratio of fast neutron component to gamma ray component to thermal neutron is less than the limiting value of IAEA recommendation. Subsequently, the depth distribution of the equivalent doses in the Snyder head phantom is calculated to evaluate the treatment characteristic. The simulation results show that the therapy rate of the beam shaping assembly based on the sandwich Fluental-LiF configuration is basically equal to that of the Fluental configuration and better than that of the LiF configuration, and the therapy time is less than that of the Fluental configuration. This means that the beam shaping assembly based on the sandwich Fluental-LiF configuration is one of the suitable options for our accelerator-based BNCT.

Research and development of various accelerator-based irradiation systems for boron neutron capture therapy (BNCT) is underway throughout the world. Many of these systems are nearing or have started clinical trials. Before the start of treatment with BNCT, the relative biological effectiveness (RBE) for the fast neutrons (over 10 keV) incident to the irradiation field must be estimated. Measurements of RBE are typically performed by biological experiments with a phantom. Although the dose deposition due to secondary gamma rays is dominant, the relative contributions of thermal neutrons (below 0.5 eV) and fast neutrons are virtually equivalent under typical irradiation conditions in a water and/or acrylic phantom. Uniform contributions to the dose deposited from thermal and fast neutrons are based in part on relatively inaccurate dose information for fast neutrons. This study sought to improve the accuracy in the dose estimation for fast neutrons by using two phantoms made of different materials in which the dose components can be separated according to differences in the interaction cross sections. The development of a "dual phantom technique" for measuring the fast neutron component of dose is reported. One phantom was filled with pure water. The other phantom was filled with a water solution of lithium hydroxide (LiOH) capitalizing on the absorbing characteristics of lithium-6 (Li-6) for thermal neutrons. Monte Carlo simulations were used to determine the ideal mixing ratio of Li-6 in LiOH solution. Changes in the depth dose distributions for each respective dose component along the central beam axis were used to assess the LiOH concentration at the 0, 0.001, 0.01, 0.1, 1, and 10 wt. % levels. Simulations were also performed with the phantom filled with 10 wt. % 6LiOH solution for 95%-enriched Li-6. A phantom was constructed containing 10 wt. % 6LiOH solution based on the simulation results. Experimental characterization of the depth dose distributions of the neutron and gamma-ray components along the central axis was performed at Heavy Water Neutron Irradiation Facility installed at Kyoto University Reactor using activation foils and thermoluminescent dosimeters, respectively. Simulation results demonstrated that the absorbing effect for thermal neutrons occurred when the LiOH concentration was over 1%. The most effective Li-6 concentration was determined to be enriched 6LiOH with a solubility approaching its upper limit. Experiments confirmed that the thermal neutron flux and secondary gamma-ray dose rate decreased substantially; however, the fast neutron flux and primary gamma-ray dose rate were hardly affected in the 10%-6LiOH phantom. It was confirmed that the dose contribution of fast neutrons is improved from approximately 10% in the pure water phantom to approximately 50% in the 10%-6LiOH phantom. The dual phantom technique using the combination of a pure water phantom and a 10%-6LiOH phantom developed in this work provides an effective method for dose estimation of the fast neutron component in BNCT. Improvement in the accuracy achieved with the proposed technique results in improved RBE estimation for biological experiments and clinical practice.

Successful implementation of Boron Neutron Capture Therapy (BNCT), requires a knowledge of the radiation dose distributions in the tissue being treated. In order to plan a BNCT treatment, a treatment planning system is necessary, which is accurate, as well as, fast enough to calculate dose distributions in minutes. Most BNCT treatment planning has focused on Monte Carlo1,2 or Discrete Ordinates3 techniques. Removal-diffusion theory may provide an alternative calculation technique that may be both accurate enough and fast enough for BNCT treatment planning. This work investigates the use of removal-diffusion theory for BNCT to calculate neutron flux distributions and, subsequently, neutron absorbed dose distributions, within a patient’s head.

Boron Neutron Capture Therapy (BNCT) is an effective and promising treatment of tumour types which are resistant to conventional therapies. The characteristics of boron neutron capture therapy (BNCT) for cancer treatment demand, in addition to sufficient fluxes of epithermal neutrons, proper conditions of the neutron sources—compact layout, flexible operation, compatibility with hospital setting, etc. The lack of proper neutron sources that can be integrable to the infrastructure of hospital or clinical facilities is a major problem. Compact neutron generators (CNGs), which are the most compact and least expensive, were a potential, alternative, solution to existing BNCT treatment facilities based on nuclear reactors. This paper will provide information about the latest CNGs technology development that has contributed to the Boron Neutron Capture Therapy (BNCT) technology improvement.

Objective To reconstruct 16-bit images of metal implants using the extended function of computed tomography (CT) imaging, and to analyze the effect of the metal CT value on calculation of dose distribution by evaluation of metal CT values in different scanning conditions. Methods A stainless steel rod and a titanium rod were inserted in a phantom. The 12-and 16-bit images and CT value distribution of metal implants were obtained by scanning the phantom using 120 kV tube voltage and 230 mA tube current. The 16-bit images and CT value distribution of metal implants were obtained by scanning the phantom using fixed tube current (230 mA) with varied tube voltage (100, 120, and 140 kV) or fixed tube voltage (120 kV) with varied tube current (180, 230, and 280 mA). In the Varian treatment planning system, a single-field plan and a parallel-opposed field plan were designed based on the CT images. The dose distribution was calculated and compared by the paired t test. Results The CT values of the stainless steel rod and the titanium rod were both 3071 HU in 12-bit CT images. In 16-bit CT images; however, the CT value of the stainless steel rod was significantly larger than that of the titanium rod. There were no significant differences in CT value of 16-bit image and dose distribution in radiotherapy plan between three scanning conditions with different tube currents. Under three scanning conditions with different tube voltages, the maximum CT values were 13568, 13127, and 12295 HU for the stainless steel rod and 8420, 7140, and 6310 HU for the titanium rod, respectively. Conclusions High-density metal implants cannot be distinguished by 12-bit images, while the distribution of metal CT value can be obtained by 16-bit images. The dose distribution of metal implants based on 12-bit images is different from that based on 16-bit images. Changes in tube voltage cause substantial changes in the CT value for metal implants, leading to changes in dose distribution in radiotherapy. Variation of tube current within a certain range causes slight changes in metal CT value and dose distribution. Key words: Metallic implants; Tomography; X-ray computed; Scanning condition; Radiotherapy dose distribution