A Monte Carlo dosimetry-based evaluation of the 7Li(p,n)7Be reaction near threshold for accelerator boron neutron capture therapy

Efficient optimization of an accelerator neutron source for neutron capture therapy using genetic algorithms

BACKGROUND

In recent years, genetic algorithms have been applied in the field of nuclear technology design, producing superior optimization results compared to traditional methods. They can be employed in the design and optimization of beam shaping assemblies (BSA) BSA to obtain the desired neutron beams. But it should be noted that the direct combination of Monte Carlo methods with genetic algorithms requires a significant amount of computational resources and time.

PURPOSE

Design and optimize BSA more efficiently to achieve neutron beams that meet specified recommendations.

METHODS

We propose an approach of NSGA II with crucial variables which are identified by multivariate statistical techniques. This approach significantly reduces the problem sizes, thus reducing the time required for optimization. We illustrate this methodology using the example of BSA design for AB-BNCT.

RESULTS

The computational efficiency has tripled with crucial variables. By using NSGA II, we obtained optimized models conforming to both the new and old version IAEA BNCT guidelines through a single optimization process and subjected them to phantom analysis. The results demonstrate that models obtained through this method can meet the IAEA recommendations with deep advantage depth (AD) and high absorbed ratio (AR).

CONCLUSION

The genetic algorithm with crucial variables displays tremendous potential in addressing BSA optimization challenges.

Measurement of energy and directional distribution of neutron ambient dose equivalent for the 7Li(p,n)7Be reaction

Double differential neutron fluence distributions were measured in the 7Li(p,n)7Be reaction for proton beam energies 7, 9 and 12 MeV. Seven liquid scintillator based detectors were employed to measure neutron fluence distributions using the Time of Flight technique. Neutron ambient dose equivalents were determined from the measured fluence distribution using ICRP (International Commission on Radiological Protection) recommended fluence to dose equivalent conversion coefficients. Neutron dose equivalents were also measured using a conventional BF3 detector based REM counter. Ambient dose equivalent measured by the REM counter is found to be in agreement with that determined from the neutron fluence spectra within their uncertainties. Angular distributions of the ambient dose equivalents were also determined from the measured fluence distributions at different angles.

Neutron flux evaluation model provided in the accelerator-based boron neutron capture therapy system employing a solid-state lithium target

An accelerator-based boron neutron capture therapy (BNCT) system employing a solid-state Li target can achieve sufficient neutron flux for treatment although the neutron flux is reduced over the lifetime of its target. In this study, the reduction was examined in the five targets, and a model was then established to represent the neutron flux. In each target, a reduction in neutron flux was observed based on the integrated proton charge on the target, and its reduction reached 28% after the integrated proton charge of 2.52 × 10⁶ mC was delivered to the target in the system. The calculated neutron flux acquired by the model was compared to the measured neutron flux based on an integrated proton charge, and the mean discrepancies were less than 0.1% in all the targets investigated. These discrepancies were comparable among the five targets examined. Thus, the reduction of the neutron flux can be represented by the model. Additionally, by adequately revising the model, it may be applicable to other BNCT systems employing a Li target, thus furthering research in this direction. Therefore, the established model will play an important role in the accelerator-based BNCT system with a solid-state Li target in controlling neutron delivery and understanding the neutron output characteristics.

Depositing a Titanium Coating on the Lithium Neutron Production Target by Magnetron Sputtering Technology

Lithium (Li) is one of the commonly used target materials for compact accelerator-based neutron source (CANS) to generate neutrons by ⁷Li(p, n)7 Be reaction. To avoid neutron yield decline caused by lithium target reacting with the air, a titanium (Ti) coating was deposited on the lithium target by magnetron sputtering technology. The color change processes of coated and bare lithium samples in the air were observed and compared to infer the chemical state of lithium qualitatively. The surface topography, thickness, and element distribution of the coating were characterized by SEM, EDS and XPS. The compositions of samples were inferred by their XRD patterns. It was found that a Ti coating with a thickness of about 200 nanometers could effectively isolate lithium from air and stabilize its chemical state in the atmosphere for at least nine hours. The Monte Carlo simulations were performed to estimate the effects of the Ti coating on the incident protons and the neutron yield. It turned out that these effects could be ignored. This research indicates that depositing a thin, titanium coating on the lithium target is feasible and effective to keep it from compounds' formation when it is exposed to the air in a short period. Such a target can be installed and replaced on an accelerator beam line in the air directly.

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.

A bi-tapered and air-gapped beam shaping assembly used for AB-BNCT

The ⁷Li(p,n)⁷Be reaction, which leads to a soft neutron field, is often chosen as the neutron producing reaction used for accelerator-based boron neutron capture therapy (AB-BNCT). This study aims to design a compact beam shaping assembly (BSA) and auxiliary system for a ⁷Li(p,n)⁷Be reaction-based neutron source and to evaluate the relationship between the BSA design and the consequent neutron beam quality for further optimization. In this study, five types of moderator shapes for the BSA model were designed. Both the in-air and in-phantom figures of merit were considered to evaluate the performance of the BSA designs. It was found that the BSA with a bi-tapered and air-gapped design could generate a high-intensity epithermal neutron beam, which could be used to treat deep-seated brain tumors within a reasonable time.

Characterization of the relationship between neutron production and thermal load on a target material in an accelerator-based boron neutron capture therapy system employing a solid-state Li target

An accelerator-based boron neutron capture therapy (BNCT) system that employs a solid-state Li target can achieve sufficient neutron flux derived from the ⁷Li(p,n) reaction. However, neutron production is complicated by the large thermal load expected on the target. The relationship between neutron production and thermal load was examined under various conditions. A target structure for neutron production consists of a Li target and a target basement. Four proton beam profiles were examined to vary the local thermal load on the target structure while maintaining a constant total thermal load. The efficiency of neutron production was evaluated with respect to the total number of protons delivered to the target structure. The target structure was also evaluated by observing its surface after certain numbers of protons were delivered. The yield of the sputtering effect was calculated via a Monte Carlo simulation to investigate whether it caused complications in neutron production. The efficiency of neutron production and the amount of damage done depended on the proton profile. A more focused proton profile resulted in greater damage. The efficiency decreased as the total number of protons delivered to the target structure increased, and the rate of decrease depended on the proton profile. The sputtering effect was not sufficiently large to be a main factor in the reduction in neutron production. The proton beam profile on the target structure was found to be important to the stable operation of the system with a solid-state Li target. The main factor in the rate of reduction in neutron production was found to be the local thermal load induced by proton irradiation of the target.

Dependence of neutrons generated by 7Li(p,n) reaction on Li thickness under free-air condition in accelerator-based boron neutron capture therapy system employing solid-state Li target

PURPOSE

An accelerator-based boron neutron capture therapy (BNCT) system with a solid-state Li target is reported to have degradation of the Li target. The degradation reduces the Li thickness, which may change spectra of the generated neutrons corresponding to the Li thickness. This study aims to examine the relationship between the Li thickness and the generated neutrons and to investigate the effects of the Li thickness on the absorbed dose in BNCT.

METHOD

The neutron energy spectra were calculated via Monte Carlo simulation for Li thicknesses ranging from 20 to 150 μm. Using the system, the saturated radioactivity of gold induced by reactions between ¹⁹⁷Au and the generated neutrons was evaluated with the simulation and the measurement, and those were compared. Additionally, for each Li thickness, the saturated radioactivity was compared with the number of generated neutrons. The absorbed doses delivered by ¹⁰B(n,α)⁷Li, ¹⁴N(n,p)¹⁴C, ¹H(n, g)²H, and (n,n') reactions in water were also calculated for each Li thickness.

RESULTS

The measurement and simulation indicated a reduction in the number of neutrons due to the degradation of the Li target. However, the absorbed doses were comparable for each Li thickness when the requisite number of neutrons for BNCT was delivered. Additionally, the saturated radioactivity of ¹⁹⁸Au could be a surrogate for the number of neutrons even if the Li thickness was varied.

CONCLUSIONS

No notable effect to the absorbed dose was observed when required neutron fluence was delivered in the BNCT even if the degradation of the Li was observed.

Beam shaping assembly design of 7Li(p,n)7Be neutron source for boron neutron capture therapy of deep-seated tumor

The development of a medical facility for boron neutron capture therapy at Budker Institute of Nuclear Physics is under way. The neutron source is based on a tandem accelerator with vacuum insulation and lithium target. The proposed accelerator is conceived to deliver a proton beam around 10 mA at 2.3 MeV proton beam. To deliver a therapeutic beam for treatment of deep-seated tumors a typical Beam Shaping Assembly (BSA) based on the source specifications has been explored. In this article, an optimized BSA based on the ⁷Li(p,n)⁷Be neutron production reaction is proposed. To evaluate the performance of the designed beam in a phantom, the parameters and the dose profiles in tissues due to the irradiation have been considered. In the simulations, we considered a proton energy of 2.3 MeV, a current of 10 mA, and boron concentrations in tumor, healthy tissues and skin of 52.5 ppm, 15 ppm and 22.5 ppm, respectively. It is found that, for a maximum punctual healthy tissue dose seated to 11 RBE-Gy, a mean dose of 56.5 RBE Gy with a minimum of 52.2 RBE Gy can be delivered to a tumor in 40 min, where the therapeutic ratio is estimated to 5.38. All of these calculations were carried out using the Monte Carlo MCNP code.

Modeling the detection efficiency of an HP-Ge detector for use in boron neutron capture therapy

The multi-foil method is commonly used to determine upon an energy spectrum of neutrons in boron neutron capture therapy. The method requires to measure the radioactivation of the foils. This study develops a simple modeling procedure of a high-purity Ge detector, which is used to measure the radioactivation, in order to calculate the detection efficiency with GEANT4. By changing four parameters from their manufacturing specifications of the detector, the simulated detection efficiency is able to reproduce the actual detection efficiency.

Evaluation of radioactivity in the bodies of mice induced by neutron exposure from an epi-thermal neutron source of an accelerator-based boron neutron capture therapy system

This study aimed to evaluate the residual radioactivity in mice induced by neutron irradiation with an accelerator-based boron neutron capture therapy (BNCT) system using a solid Li target. The radionuclides and their activities were evaluated using a high-purity germanium (HP-Ge) detector. The saturated radioactivity of the irradiated mouse was estimated to assess the radiation protection needs for using the accelerator-based BNCT system. ²⁴Na, ³⁸Cl, ^80mBr, ⁸²Br, ⁵⁶Mn, and ⁴²K were identified, and their saturated radioactivities were (1.4 ± 0.1) × 10², (2.2 ± 0.1) × 10¹, (3.4 ± 0.4) × 10², 2.8 ± 0.1, 8.0 ± 0.1, and (3.8 ± 0.1) × 10¹ Bq/g/mA, respectively. The ²⁴Na activation rate at a given neutron fluence was found to be consistent with the value reported from nuclear-reactor-based BNCT experiments. The induced activity of each nuclide can be estimated by entering the saturated activity of each nuclide, sample mass, irradiation time, and proton current into the derived activation equation in our accelerator-based BNCT system.

COMPREHENSIVE RADIATION DOSE MEASUREMENTS AND MONTE CARLO SIMULATION FOR THE 7Li(p,n) ACCELERATOR NEUTRON FIELD

In order to investigate the radiation dose dependence on the incident proton energy, neutron and gamma-ray doses were measured using a tissue-equivalent proportional counter in the proton energy range of 1.95-2.50 MeV for the McMaster ⁷Li(p,n) neutron facility. Microdosimetric spectra were collected, and absorbed doses were determined at various positions inside the irradiation cavity, along the lateral axis and outside the shield to find out the spatial distributions of neutron and gamma-ray doses for each proton energy. In parallel with the absorbed dose measurements, MCNP Monte Carlo simulations were carried out and neutron fluence spectra were computed at various positions, which enabled determination of the neutron weighting factors. It was found that neutrons make a substantially dominant contribution to the total equivalent dose for most proton energies and positions. The effective dose for a human subject increased from 0.058 to 1.306 μSv μA^-1 min^-1 with the increase of proton energy from 1.95 to 2.5 MeV. It is expected that the reported data will be useful for ⁷Li(p,n) accelerator neutron users.

Triple ionization chamber method for clinical dose monitoring with a Be-covered Li BNCT field

PURPOSE

Fast neutron, gamma-ray, and boron doses have different relative biological effectiveness (RBE). In boron neutron capture therapy (BNCT), the clinical dose is the total of these dose components multiplied by their RBE. Clinical dose monitoring is necessary for quality assurance of the irradiation profile; therefore, the fast neutron, gamma-ray, and boron doses should be separately monitored. To estimate these doses separately, and to monitor the boron dose without monitoring the thermal neutron fluence, the authors propose a triple ionization chamber method using graphite-walled carbon dioxide gas (C-CO₂), tissue-equivalent plastic-walled tissue-equivalent gas (TE-TE), and boron-loaded tissue-equivalent plastic-walled tissue-equivalent gas [TE(B)-TE] chambers. To use this method for dose monitoring for a neutron and gamma-ray field moderated by D₂O from a Be-covered Li target (Be-covered Li BNCT field), the relative sensitivities of these ionization chambers are required. The relative sensitivities of the TE-TE, C-CO₂, and TE(B)-TE chambers to fast neutron, gamma-ray, and boron doses are calculated with the particle and heavy-ion transport code system (PHITS).

METHODS

The relative sensitivity of the TE(B)-TE chamber is calculated with the same method as for the TE-TE and C-CO₂ chambers in the paired chamber method. In the Be-covered Li BNCT field, the relative sensitivities of the ionization chambers to fast neutron, gamma-ray, and boron doses are calculated from the kerma ratios, mass attenuation coefficient tissue-to-wall ratios, and W-values. The Be-covered Li BNCT field consists of neutrons and gamma-rays which are emitted from a Be-covered Li target, and this resultant field is simulated by using PHITS with the cross section library of ENDF-VII. The kerma ratios and mass attenuation coefficient tissue-to-wall ratios are determined from the energy spectra of neutrons and gamma-rays in the Be-covered Li BNCT field. The W-value is calculated from recoil charged particle spectra by the collision of neutrons and gamma-rays with the wall and gas materials of the ionization chambers in the gas cavities of TE-TE, C-CO₂, and TE(B)-TE chambers (¹⁰B concentrations of 10, 50, and 100 ppm in the TE-wall).

RESULTS

The calculated relative sensitivity of the C-CO₂ chamber to the fast neutron dose in the Be-covered Li BNCT field is 0.029, and those of the TE-TE and TE(B)-TE chambers are both equal to 0.965. The relative sensitivities of the C-CO₂, TE-TE, and TE(B)-TE chambers to the gamma-ray dose in the Be-covered Li BNCT field are all 1 within the 1% calculation uncertainty. The relative sensitivities of TE(B)-TE to boron dose with concentrations of 10, 50, and 100 ppm ¹⁰B are calculated to be 0.865 times the ratio of the in-tumor to in-chamber wall boron concentration.

CONCLUSIONS

The fast neutron, gamma-ray, and boron doses of a tumor in-air can be separately monitored by the triple ionization chamber method in the Be-covered Li BNCT field. The results show that these doses can be easily converted to the clinical dose with the depth correction factor in the body and the RBE.

Application of an ultraminiature thermal neutron monitor for irradiation field study of accelerator-based neutron capture therapy

Phantom experiments to evaluate thermal neutron flux distribution were performed using the Scintillator with Optical Fiber (SOF) detector, which was developed as a thermal neutron monitor during boron neutron capture therapy (BNCT) irradiation. Compared with the gold wire activation method and Monte Carlo N-particle (MCNP) calculations, it was confirmed that the SOF detector is capable of measuring thermal neutron flux as low as 10(5) n/cm(2)/s with sufficient accuracy. The SOF detector will be useful for phantom experiments with BNCT neutron fields from low-current accelerator-based neutron sources.

Dosimetric performance evaluation regarding proton beam incident angles of a lithium-based AB-BNCT design

The (7)Li(p,xn)(7)Be nuclear reaction, based on the low-energy protons, could produce soft neutrons for accelerator-based boron neutron capture therapy (AB-BNCT). Based on the fact that the induced neutron field is relatively divergent, the relationship between the incident angle of proton beam and the neutron beam quality was evaluated in this study. To provide an intense epithermal neutron beam, a beam-shaping assembly (BSA) was designed. And a modified Snyder head phantom was used in the calculations for evaluating the dosimetric performance. From the calculated results, the intensity of epithermal neutrons increased with the increase in proton incident angle. Hence, either the irradiation time or the required proton current can be reduced. When the incident angle of 2.5-MeV proton beam is 120°, the required proton current is ∼13.3 mA for an irradiation time of half an hour.

A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft tissue). This code can transport protons in wide range of energies (up to 200 MeV for proton). The validity of the fast Monte Carlo (MC) code is evaluated with data MCNPX as a reference code. While analytical pencil beam algorithm transport shows great errors (up to 10%) near small high density heterogeneities, there was less than 2% deviation of MCNPX results in our dose calculation and isodose distribution. In terms of speed, the code runs 200 times faster than MCNPX. In the Fast MC code which is developed in this work, it takes the system less than 2 minutes to calculate dose for 10(6) particles in an Intel Core 2 Duo 2.66 GHZ desktop computer.

Revisiting the (7)Li(p,n)(7)Be reaction near threshold

In this work we review all the available experimental neutron data for the (7)Li(p,n) reaction near threshold which is necessary to obtain an accurate source model for Monte Carlo simulations in Boron Neutron Capture Therapy. Scattered published experimental results such as cross sections, differential neutron yields and total yields were collected and analyzed, exploring the sensitivity of the fitting parameters to the different possible variables and deriving a consistent working set of parameters to evaluate the neutron source near threshold.

Note: Proton irradiation at kilowatt-power and neutron production from a free-surface liquid-lithium target

The free-surface Liquid-Lithium Target, recently developed at Soreq Applied Research Accelerator Facility (SARAF), was successfully used with a 1.9 MeV, 1.2 mA (2.3 kW) continuous-wave proton beam. Neutrons (~2 × 10(10) n/s having a peak energy of ~27 keV) from the (7)Li(p,n)(7)Be reaction were detected with a fission-chamber detector and by gold activation targets positioned in the forward direction. The setup is being used for nuclear astrophysics experiments to study neutron-induced reactions at stellar energies and to demonstrate the feasibility of accelerator-based boron neutron capture therapy.

Study on detecting spatial distribution of neutrons and gamma rays using a multi-imaging plate system

In order to measure the spatial distributions of neutrons and gamma rays separately using the imaging plate, the requirement for the converter to enhance specific component was investigated with the PHITS code. Consequently, enhancing fast neutrons using recoil protons from epoxy resin was not effective due to high sensitivity of the imaging plate to gamma rays. However, the converter of epoxy resin doped with (10)B was found to have potential for thermal and epithermal neutrons, and graphite for gamma rays.

A TPD and AR based comparison of accelerator neutron irradiation fields between (7)Li and W targets for BNCT

The characteristics of moderator assembly dimension was investigated for the usage of (7)Li(p,n) neutrons by 2.3-2.8MeV protons and W(p,n) neutrons by 50MeV protons. The indexes were the treatable protocol depth (TPD) and advantage depth (AD). Consequently, a configuration for W target with the Fe filter, Fluental moderator, Pb reflector showed the TPD of 5.8cm and AD of 9.3cm. Comparable indexes were found for the Li target in a geometry with the MgF2 moderator and Teflon reflector.

High-power liquid-lithium jet target for neutron production

A compact liquid-lithium target (LiLiT) was built and tested with a high-power electron gun at the Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the (7)Li(p,n)(7)Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy. The liquid-lithium jet target acts both as neutron-producing target and beam dump by removing the beam thermal power (>5 kW, >1 MW/cm(3)) with fast transport. The target was designed based on a thermal model, accompanied by a detailed calculation of the (7)Li(p,n) neutron yield, energy distribution, and angular distribution. Liquid lithium is circulated through the target loop at ~200 °C and generates a stable 1.5 mm-thick film flowing at a velocity up to 7 m/s onto a concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power areal densities of >4 kW/cm(2) and volume power density of ~2 MW/cm(3) at a lithium flow of ~4 m/s while maintaining stable temperature and vacuum conditions. The LiLiT setup is presently in online commissioning stage for high-intensity proton beam irradiation (1.91-2.5 MeV, 1-2 mA) at SARAF.

An accelerator-based neutron microbeam system for studies of radiation effects

A novel neutron microbeam is being developed at the Radiological Research Accelerator Facility (RARAF) of Columbia University. The RARAF microbeam facility has been used for studies of radiation bystander effects in mammalian cells for many years. Now a prototype neutron microbeam is being developed that can be used for bystander effect studies. The neutron microbeam design here is based on the existing charged particle microbeam technology at the RARAF. The principle of the neutron microbeam is to use the proton beam with a micrometre-sized diameter impinging on a very thin lithium fluoride target system. From the kinematics of the ⁷Li(p,n)⁷Be reaction near the threshold of 1.881 MeV, the neutron beam is confined within a narrow, forward solid angle. Calculations show that the neutron spot using a target with a 17-µm thick gold backing foil will be <20 µm in diameter for cells attached to a 3.8-µm thick propylene-bottomed cell dish in contact with the target backing. The neutron flux will roughly be 2000 per second based on the current beam setup at the RARAF singleton accelerator. The dose rate will be about 200 mGy min⁻¹. The principle of this neutron microbeam system has been preliminarily tested at the RARAF using a collimated proton beam. The imaging of the neutron beam was performed using novel fluorescent nuclear track detector technology based on Mg-doped luminescent aluminum oxide single crystals and confocal laser scanning fluorescent microscopy.

Development of monitoring method of spatial neutron distribution in neutrons-gamma rays mixed field using imaging plate for NCT--depression of the field

The degree of depression in the neutron field caused by neutron absorption in the materials of an imaging plate (IP) was investigated using MCNP-4C. Consequently, the IP doped with Gd, which reproduced the distribution of (157)Gd(n,γ)(158)Gd reaction rate in the previous study, depresses the relative distribution by about 50%. The depression for the IP in which Gd is replaced with similar amount of B atoms was estimated to be about 10%. The signal intensity for this IP is estimated to be at a similar level with that for Gd-doped IP.

High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy

A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks.

Dose calculation from a D-D-reaction-based BSA for boron neutron capture synovectomy

Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D-D-reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values.

Characteristics of proton beam scanning dependent on Li target thickness from the viewpoint of heat removal and material strength for accelerator-based BNCT

This study demonstrates the characterization of proton spot scanning on a Li target assembly for accelerator-based BNCT from the viewpoint of heat removal and material strength. These characteristics are investigated as to their dependence on the Li target thickness, considering that the Cu backing plate has more suitable heat removal properties than Li. Two situations are considered in this paper, i.e. the cyclic operation of the spot scanning, and a stalled spot scanning cycle where the proton beam stays focused on a single position on the Li target. It was found that the maximum of the Li temperature and the strain of the Cu backing increase as the cycle period increases. A cycle period less than 120 ms (over 8.3 Hz of frequency) enables the Li temperature to be kept below 150 degrees C and a cycle of less than 115 ms (8.7 Hz) keeps the Cu strain below the critical value for a 230 microm thick Li target, though the values are evaluated conservatively. Against expectation, the Li temperature and Cu strain are larger for a 100 microm thick target than for a 230 microm target. The required cycle period in this case is 23 ms (43 Hz) for maintaining a reasonable Li temperature and 9 ms (110 Hz) to prevent Cu fatigue fracture. For a stall in the spot scanning cycle, the Cu temperature increases as the beam shutdown time increases. The time for Cu to reach its melting point is estimated to be 4.2 ms at the surface, 20 ms at 1mm depth, for both of 100 and 230 microm thick targets. At least 34 ms is estimated to be enough to make a hole on Cu backing plate. A beam shutdown mechanism with a response time of about 20 ms is therefore required.

Variations in lithium target thickness and proton energy stability for the near-threshold 7Li(p,n)7Be accelerator-based BNCT

The usable range of thickness for the solid lithium target in the accelerator-based neutron production for BNCT via the near-threshold (7)Li(p,n)(7)Be reaction was investigated. While the feasibility of using a (7)Li-target with thickness equal to that which is required to slow down a mono-energetic 1.900 MeV incident proton to the 1.881 MeV threshold of the (7)Li(p,n)(7)Be reaction (i.e., t(min) = 2.33 microm) has already been demonstrated, dosimetric properties of neutron fields from targets greater than t(min) were assessed as thicker targets would last longer and offer more stable neutron production. Additionally, the characteristics of neutron fields generated by (7)Li(p,n)(7)Be for Gaussian incident protons with mean energy of 1.900 MeV were evaluated at a (7)Li-target thickness t(min). The main evaluation index applied in this study was the treatable protocol depth (TPD) which corresponds to the depth in an irradiated medium that satisfies the requirements of the adapted dose protocol. A maximum TPD (TPD(max)) was obtained for each irradiation condition from the relationship between the TPD and the thickness of boron dose enhancer (BDE) used. For a mono-energetic 1.900 MeV proton beam, the deepest TPD(max) of 3.88 cm was attained at the (7)Li-target thickness of t(min) and a polyethylene BDE of 1.10 cm. When the intended TPD for a BNCT clinical treatment is shallower than the deepest TPD(max), the usable (7)Li-target thickness would be between t(min) and an upper limit t(upper) whose value depends on the BDE thickness used. In terms of the effect of stability of the incident proton energy, Gaussian incident proton energies stable to within +/-10 keV of 1.900 MeV were found to be feasible for the neutron production via the near-threshold (7)Li(p,n)(7)Be reaction for BNCT provided that a suitable BDE is used.

Characterization of moderator assembly dimension for accelerator boron neutron capture therapy of brain tumors using 7Li(p, n) neutrons at proton energy of 2.5 MeV

The characteristics of moderator assembly dimension are investigated for the usage of 7Li(p,n) neutrons by 2.5 MeV protons in boron newtron capture therapy (BNCT) of brain tumors in the present study. The indexes checked are treatable protocol depth (TPD), which is the greatest depth of the region satisfying the dose requirements in BNCT protocol, proton current necessary to complete BNCT by 1 h irradiation, and the heat flux deposited in the Li target which should be removed. Assumed materials are D2O for moderator, and mixture of polyethylene and LiF with 50 wt % for collimator. Dose distributions have been computed with MCNP 4B and 4C codes. Consequently, realized TPD does not show a monotonical tendency for the Li target diameter. However, the necessary proton current and heat flux in the Li target decreases as the Li target diameter increases, while this trend reverses at around 10 cm of the Li target diameter for the necessary proton current in the condition of this study. As to the moderator diameter, TPD does not exhibit an apparent dependence. On the other hand, necessary proton current and heat flux decrease as the moderator diameter increases, and this tendency saturates at around 60 cm of the moderator diameter in this study. As to the collimator, increase in inner diameter is suitable from the viewpoint of increasing TPD and decreasing necessary proton current and heat flux, while these indexes do not show apparent difference for collimator inner diameters over 14 cm for the parameters treated here. The practical viewpoint in selecting the parameters of moderator assembly dimension is to increase TPD, within the technically possible condition of accelerated proton current and heat removal from the Li target. In this process, the values for which the resultant characteristics mentioned above saturate or reverse would be important factors.

TPD-based evaluation of near threshold mono-energetic proton energies for the7Li(p,n)7Be production of neutrons for BNCT

An evaluation of mono-energetic proton energies ranging from 1.885 MeV to 1.920 MeV was carried out to determine the viability of these near threshold energies in producing neutrons for BNCT via the (7)Li(p,n)(7)Be reaction. Neutron fields generated at these proton energies were assessed using the treatable protocol depth (TPD) and the maximum TPD (TPD(max)) as evaluation indices. The heavy charged particle (HCP) dose rate to tumour was likewise applied as a figure of merit in order to account for irradiation time and required proton current. Incident proton energies closer to the reaction threshold generated deeper TPDs compared to higher energy protons when no boron dose enhancers (BDE) were placed in the irradiation field. Introducing a BDE resulted in improved TPDs for high proton energies but their achievable TPD(max) were comparatively lower than that obtained for lower proton energies. In terms of the HCP dose rate to tumour, higher proton energies generated neutron fields that yielded higher dose rates both at TPD(max) and at fixed depths of comparison. This infers that higher currents are required to deliver the prescribed treatment dose to tumours for proton beams with energies closer to the (7)Li(p,n)(7)Be reaction threshold and more achievable proton currents of around 10 mA or less for proton energies from 1.900 MeV and above. The dependence on incident proton energy of the TPD, TPD(max) and the HCP dose rate to tumour with respect to the (10)B concentration in tumour and healthy tissues were also clarified in this study. Increasing the (10)B concentration in tumour while maintaining a constant T/N ratio resulted in deeper TPD(max) where a greater change in TPD(max) was obtained for proton energies closer to the (7)Li(p,n)(7)Be reaction threshold. The HCP dose rates to tumour for all proton energies also went up, with the higher proton energies benefiting more from the increased (10)B concentration.

Monte Carlo simulation of neutron irradiation facility developed for accelerator based in vivo neutron activation measurements in human hand bones

The neutron irradiation facility developed at the McMaster University 3 MV Van de Graaff accelerator was employed to assess in vivo elemental content of aluminum and manganese in human hands. These measurements were carried out to monitor the long-term exposure of these potentially toxic trace elements through hand bone levels. The dose equivalent delivered to a patient during irradiation procedure is the limiting factor for IVNAA measurements. This article describes a method to estimate the average radiation dose equivalent delivered to the patient's hand during irradiation. The computational method described in this work augments the dose measurements carried out earlier [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724]. This method employs the Monte Carlo simulation of hand irradiation facility using MCNP4B. Based on the estimated dose equivalents received by the patient hand, the proposed irradiation procedure for the IVNAA measurement of manganese in human hands [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724] with normal (1 ppm) and elevated manganese content can be carried out with a reasonably low dose of 31 mSv to the hand. Sixty-three percent of the total dose equivalent is delivered by non-useful fast group (> 10 keV); the filtration of this neutron group from the beam will further decrease the dose equivalent to the patient's hand.

Characteristics of boron-dose enhancer dependent on dose protocol and 10B concentration for BNCT using near-threshold 7Li(p,n)7Be direct neutrons

The dependence of boron-dose enhancer (BDE) characteristics on dose protocol and 10B concentration was evaluated for BNCT using near-threshold 7Li(p,n)7Be direct neutrons. The treatable protocol depth (TPD) was utilized as an evaluation index. MCNP calculations were performed for near-threshold 7Li(p,n)7Be at a proton energy of 1.900 MeV and for a polyethylene BDE. The effect of dose protocol on BDE characteristics was reflected in terms of the optimum BDE thickness needed for maximum TPD which was found to be independent of the treatable dose but was observed to vary for different combinations of the tolerance doses for heavy charged particles and gamma rays. For the 10B concentration dependence, the TPD was increased by increasing the T/N ratio, i.e., the ratio of the 10B concentration in the tumour (10B(Tumour)) to that in the normal tissue (10B(Normal)), and by increasing 10B(Tumour) and 10B(Normal) at constant T/N ratio. It was found that the use of BDE becomes unnecessary from the viewpoint of increasing the TPD, when 10B(Tumour) is over a certain level which is decided by the conditions of the dose protocol.

Characteristics of BDE dependent on 10B concentration for accelerator-based BNCT using near-threshold 7Li(p,n)7Be direct neutrons

The characteristics boron-dose enhancer (BDE) was evaluated as to the dependence on the (10)B concentration for BNCT using near-threshold (7)Li(p,n)(7)Be direct neutrons. The treatable protocol depth (TPD) was utilized as an evaluation index. MCNP-4B calculations were performed for near-threshold (7)Li(p,n)(7)Be at a proton energy of 1.900MeV and for a polyethylene BDE. Consequently, the TPD was increased by increasing T/N ratio, i.e., the ratio of the (10)B concentration in the tumor ((10)B(Tumor)) to that in the normal tissue ((10)B(Normal)), and by increasing (10)B(Tumor) and (10)B(Normal) for constant T/N ratio. It has been found that the BDE becomes unnecessary from the viewpoint of increasing the TPD, when (10)B(Tumor) is over a certain level.

Optimization parameters for BDE in BNCT using near threshold 7Li(p,n)7Be direct neutrons

The dose contribution of (10)B(n,alpha)(7)Li reaction in BNCT using near threshold (7)Li(p,n)(7)Be direct neutrons can be increased through the use of materials referred to as boron-dose enhancers (BDE). In this paper, possible BDE optimization criteria were determined from the characteristics of candidate BDE materials namely (C(2)H(4))(n), (C(2)H(3)F)(n), (C(2)H(2)F(2))(n), (C(2)HF(3))(n), (C(2)D(4))(n), (C(2)F(4))(n), beryllium metal, graphite, D(2)O and (7)LiF. The treatable protocol depth (TPD) was used as the assessment index for evaluating the effect of these materials on the dose distribution in a medium undergoing BNCT using near threshold (7)Li(p,n)(7)Be direct neutrons. The maximum TPD (TPD(max)) did not exhibit an explicit dependence on material type as evidenced by its small range and arbitrary variations. The dependence of TPD on BDE thickness was influenced by the BDE material used as indicated by the sharply peaked TPD versus BDE thickness curves for materials with hydrogen compared to the broader curves obtained for those without hydrogen. The BDE thickness required to achieve TPD(max) (BDE(TPD(max))) were also found to be thinner for materials with hydrogen. The TPD(max), the dependence of TPD on BDE thickness, and the BDE(TPD(max)) were established as appropriate BDE optimization parameters. Based on these criteria and other practical considerations, the suitable choice as BDE among the candidate materials considered in this study for treatments involving tumors located at shallow depths would be (C(2)H(4))(n) while beryllium metal was judged as more appropriate for treatment of deep-seated tumors.

Evaluation of the characteristics of boron-dose enhancer (BDE) materials for BNCT using near threshold7Li(p,n)7Be direct neutrons

The characteristics of a number of candidate boron-dose enhancer (BDE) materials for boron neutron capture therapy (BNCT) using near threshold 7Li(p,n)7Be direct neutrons were evaluated based on the treatable protocol depth (TPD), defined in this paper. Simulation calculations were carried out by means of MCNP-4B transport code for candidate BDE materials, namely, (C2H4)n, (C2H3F)n, (C2H2F2)n, (C2HF3)n, (C2D4)n, (C2F4)n, beryllium metal, graphite, D2O and 7LiF. Dose protocols applied were those used for intra-operative BNCT treatment for brain tumour currently used in Japan. The maximum TPD (TPDmax) for each BDE material was found to be between 4 cm and 5 cm in the order of (C2H4)n < (C2H3F)n < (C2H2F2)n < (C2HF3)n < beryllium metal < (C2D4)n < graphite < (C2F4)n < D2O < 7LiF. Based on the small and arbitrary variations in the TPDmax for these materials, an explicit advantage of a candidate BDE material could not be established from the TPDmax alone. The dependence of TPD on BDE thickness was found to be influenced by the type of BDE material. For materials with hydrogen, sharp variations in TPD were observed, while those without hydrogen exhibited more moderate fluctuations in TPD as the BDE thickness was varied. The BDE thickness corresponding to TPDmax (BDE(TPDmax)) was also found to depend on the type of BDE material used. Thicker BDE(TPDmax), obtained mostly for BDE materials without hydrogen, significantly reduced the dose rates within the phantom. The TPDmax, the dependence of TPD on BDE thickness and the BDE (TPDmax) were ascertained as appropriate optimization criteria in choosing suitable BDE materials for BNCT. Among the candidate BDE materials considered in this study. (C2H4)n was judged as the suitable material for near-surface tumours and beryllium metal for deeper tumours based on these optimization criteria and other practical considerations.

Feasibility study on epithermal neutron field for cyclotron-based boron neutron capture therapy

To realize the accelerator-based boron neutron capture therapy (BNCT) at the Cyclotron and Radioisotope Center of Tohoku University, the feasibility of a cyclotron-based BNCT was evaluated. This study focuses on optimizing the epithermal neutron field with an energy spectrum and intensity suitable for BNCT for various combinations of neutron-producing reactions and moderator materials. Neutrons emitted at 90 degrees from a thick (stopping-length) Ta target, bombarded by 50 MeV protons of 300 microA beam current, were selected as a neutron source, based on the measurement of angular distributions and neutron energy spectra. As assembly composed of iron, AlF3/Al/6LiF, and lead was chosen as moderators, based on the simulation trials using the MCNPX code. The depth dose distributions in a cylindrical phantom, calculated with the MCNPX code, showed that, within 1 h of therapeutic time, the best moderator assembly, which is 30-cm-thick iron, 39-cm-thick AlF3/Al/6LiF, and 1-cm-thick lead, provides an epithermal neutron flux of 0.7 x 10(9) [n cm(-2) s(-1)]. This results in a tumor dose of 20.9 Gy-eq at a depth of 8 cm in the phantom, which is 6.4 Gy-eq higher than that of the Brookhaven Medical Research Reactor at the equivalent condition of maximum normal tissue tolerance. The beam power of the cyclotron is 15 kW, which is much lower than other accelerator-based BNCT proposals.

Accelerator-based epithermal neutron sources for boron neutron capture therapy of brain tumors

This paper reviews the development of low-energy light ion accelerator-based neutron sources (ABNSs) for the treatment of brain tumors through an intact scalp and skull using boron neutron capture therapy (BNCT). A major advantage of an ABNS for BNCT over reactor-based neutron sources is the potential for siting within a hospital. Consequently, light-ion accelerators that are injectors to larger machines in high-energy physics facilities are not considered. An ABNS for BNCT is composed of: (1) the accelerator hardware for producing a high current charged particle beam, (2) an appropriate neutron-producing target and target heat removal system (HRS), and (3) a moderator/reflector assembly to render the flux energy spectrum of neutrons produced in the target suitable for patient irradiation. As a consequence of the efforts of researchers throughout the world, progress has been made on the design, manufacture, and testing of these three major components. Although an ABNS facility has not yet been built that has optimally assembled these three components, the feasibility of clinically useful ABNSs has been clearly established. Both electrostatic and radio frequency linear accelerators of reasonable cost (approximately 1.5 M dollars) appear to be capable of producing charged particle beams, with combinations of accelerated particle energy (a few MeV) and beam currents (approximately 10 mA) that are suitable for a hospital-based ABNS for BNCT. The specific accelerator performance requirements depend upon the charged particle reaction by which neutrons are produced in the target and the clinical requirements for neutron field quality and intensity. The accelerator performance requirements are more demanding for beryllium than for lithium as a target. However, beryllium targets are more easily cooled. The accelerator performance requirements are also more demanding for greater neutron field quality and intensity. Target HRSs that are based on submerged-jet impingement and the use of microchannels have emerged as viable target cooling options. Neutron fields for reactor-based neutron sources provide an obvious basis of comparison for ABNS field quality. This paper compares Monte Carlo calculations of neutron field quality for an ABNS and an idealized standard reactor neutron field (ISRNF). The comparison shows that with lithium as a target, an ABNS can create a neutron field with a field quality that is significantly better (by a factor of approximately 1.2, as judged by the relative biological effectiveness (RBE)-dose that can be delivered to a tumor at a depth of 6cm) than that for the ISRNF. Also, for a beam current of 10 mA, the treatment time is calculated to be reasonable (approximately 30 min) for the boron concentrations that have been assumed.