Accelerator-based Neutron Source Research Articles

Measurement of the 10B(d,α0)8Be, 10B(d,α1)8Be*, 10B(d,p2)9Be*, 11B(d,α0)9Be, and 11B(d,α2)9Be* reactions cross-sections at the deuteron energies up to 2.2 MeV

The deuteron-boron fusion reaction produces many different charged particles. Data on the reactions cross-sections differ among different authors, so obtaining experimental data is still relevant. Measurements of the reactions cross-section were carried out at the accelerator-based neutron source VITA at Budker Institute of Nuclear Physics (Novosibirsk, Russia) using an α-spectrometer. The 10B(d,α0)8Be, 10B(d,α1)8Be*, 10B(d,p2)9Be*, 11B(d,α0)9Be, and 11B(d,α2)9Be* reactions cross-sections at the deuteron energies up to 2.2 MeV have been measured. The obtained data are presented in tabular form.

Properties analysis on interlayer within lithium target and its beam uniformity for accelerator-based Boron Neutron Capture Therapy

In order to enhance the performance and lifetime of the lithium target used in accelerator-based neutron sources for Boron Neutron Capture Therapy (BNCT) treatment, an exploration of target design was conducted based on the 2.8 MeV, 20 mA proton beam. A comparison between scanning magnets and octupole magnets was performed for beam uniform, with octupole magnets selected to effectively avoid localized high thermal densities over short durations. Exploration was conducted on the performance of tantalum and vanadium as interlayers within the lithium target, considering aspects such as cooling, hydrogen diffusion, and neutron performances. This study revealed that, as the majority of energy deposition occurs within the interlayer, the presence of an appropriately thick tantalum or vanadium interlayer has minimal impact on cooling effectiveness, ensuring temperatures remain below 144 °C. The addition of an interlayer effectively reduces the maximum hydrogen concentration in copper, thus preventing copper blistering. Within the investigated thickness range, the interlayer does not affect neutron spectrum in the forward direction of the target, mitigating concerns regarding its impact on beam shaping.

Implementation of optically simulated luminescent dosimeter for quality control of gamma ray dose of an accelerator-based neutron source.

Neutron beams utilized for performing BNCT are composed of a mixture of neutrons and gamma rays. Although much of the dose delivered to the cancer cells comes from the high LET particles produced by the boron neutron capture reaction, the dose delivered to the healthy tissues from unwanted gamma rays cannot be ignored. With the increase in the number of accelerators for BNCT, a detector system that is capable of measuring gamma ray dose in a mixed neutron/gamma irradiation field is crucial. Currently, BeO TLDs encased in quartz glass are used to measure gamma ray dose in a BNCT irradiation field. However, this type of TLD is no longer commercially available. A replacement dosimetry system is required to perform the recommended ongoing quality assurance of gamma ray measurement for a clinical BNCT system. The purpose of this study is to evaluate the characteristics of a BeO OSLD detector system under a mixed neutron and gamma ray irradiation field and to assess the suitability of the system for routine quality assurance measurements of an accelerator-based BNCT facility. The myOSLD system by RadPro International GmbH was evaluated using the accelerator-based neutron source designed for clinical BNCT (NeuCure BNCT system). The readout constancy, linearity, dose rate effect, and fading effect of the OSLD were evaluated. Free-in-air and water phantom measurements were performed and compared with the TLD results and Monte Carlo simulation results. The PHITS Monte Carlo code was used for this study. The readout constancy was found to be stable over a month-long period and similar to the TLD results. The OSLD readout signal was found to be linear, with a high coefficient of determination (R2≥0.999) up to a proton charge of 3.6C. There was no significant signal fading or dose rate dependency. The central axis depth dose and off-axis dose profile measurements agreed with both the TLD and Monte Carlo simulation results, within one standard deviation. The myOSLD system was characterized using an accelerator system designed for clinical BNCT. The experimental measurements confirmed the OSLD achieved similar, if not superior to, the currently utilized dosimetry system for routine QA of an accelerator-based BNCT system. The OSLD system would be a suitable replacement for the current TLD system for performing routine QA of gamma ray dose measurement in a BNCT irradiation field.

New technologies for beam spectrometry, quality assurance, real-time monitoring and microdosimetry in BNCT

Boron neutron capture therapy (BNCT) is a next-generation radiotherapy that combines neutron beams with boron compounds that selectively accumulate in cancer cells. Because this therapy requires neutrons for irradiation, compact accelerator-based neutron source devices that can be installed in hospitals are being developed around the world. Clinical trials of several devices have been conducted in Japan, China, and South Korea. The most advanced device was approved by the Japanese regulatory authorities in 2020. As a result, BNCT with the device is being implemented as an insured therapy for recurrent head and neck cancer at two hospitals in Japan. Thus, the development of accelerator-based neutron generators is proceeding in the field of BNCT. However, much remains to be done to establish this therapy as a common cancer treatment and to make it available worldwide. Dosimetry methods in BNCT are one of the issues. In the treatment, the dose given to a patient by neutron irradiation has to be estimated accurately. However, it is difficult to measure neutrons accurately and in real time. Currently, the neutron dose delivered to patients during treatment is not measured. Instead, the current of the charged particles that irradiate the neutron target material is measured to indirectly evaluate the neutron fluence and dose. In addition, accurate dose estimation based on reactions with neutrons and various elements in a human body uses the Monte Carlo method, which is computationally time-consuming. To address these dosimetry issues, several neutron fluence and dose measurement methods are being developed. This review briefly describes the current dosimetry methods and then presents several methods under development that allow real-time neutron measurements applicable to BNCT.

RF electromagnetic design and field tuning study of the coupled ladder RFQ-IH DTL structure

The coupled ladder RFQ-IH DTL structure is designed for transportable accelerator-based neutron source to accelerate the proton beam to 2.5 MeV with a peak current of 15 mA. The Kombinierte Null Grad Struktur (KONUS) dynamics scheme is chosen in the IH-DTL section to obtain compact structure, high transmission efficiency and small emittance growth. Three KONUS periods, including two quadrupole doublets are housed in the IH-DTL section. The field distribution of coupled structure is studied and the evaluation of the coupled field is quantified. The optimization voltage ratio and electric flatness are carried out based on three-dimension full model. In addition, the structure error analysis is performed to evaluate the acceptance of different errors. The tuning system is designed to compensate the machining and alignment errors and the field tuning strategy is proposed for the coupled cavity. The RF electromagnetic design and optimization, structure error analysis and field tuning will be comprehensively presented in this paper.

Study of the out-of-field dose from an accelerator-based neutron source for boron neutron capture therapy

One important issue in Boron Neutron Capture Therapy is the delivered dose to the tissues outside the tumor. An international standard for light ion beam systems sets two recommended limits for out-of-field dose based on distance from the field edge: maximum absorbed dose from all radiation types shall not exceed 0.5 % of the maximum dose at distances 15 cm to 50 cm from the field edge. At distances >50 cm from the field edge, the maximum absorbed dose shall not exceed 0.1 %. This paper is a continuation of our previous works focused on the design of an accelerator-based neutron source for BNCT. We already designed a novel Beam Shape Assembly which meets the IAEA criteria for BNCT treatments. Using this BSA, in the present work, we characterize by Monte Carlo simulations the dose outside the neutron field. The out-of-field dose has been assessed via estimates using the ambient and equivalent dose. Also the boron uptake in healthy tissues has been analyzed for the equivalent dose computation. It is concluded that our design for a future accelerator-based source for BNCT meets reasonably well the criteria defined from other forms of radiotherapy on both equivalent and effective dose outside the field.

A correction algorithm of geometric unsharpness for neutron radiographs via adaptive fusion and total variation

The mobility of accelerator-based neutron sources has laid the foundation for the miniaturization of neutron radiography facilities in response to the increasing industrial requirements. Compared to traditional reactors and spallation neutron sources, miniaturized neutron radiography facilities based on accelerator neutron sources face challenges in achieving a sufficiently high collimation ratio. It leads to a pronounced level of geometric unsharpness on neutron radiographs, which arises from a finite source size and beam divergence. The presence of blurriness in radiographs significantly impairs the ability to accurately visualize the internal structures of the test objects, thus hindering observation. An algorithm combining adaptive fusion with total variation was proposed to correct the geometric unsharpness for neutron radiographs in this paper. In contrast to most traditional restoration algorithms that relied on regularization terms or filters, our proposed method offered both initial restoration and further restoration. The initial restoration procedure involved the utilization of the Richardson-Lucy algorithm coupled with the l0 smoothing method. This adaptive fusion technique was employed to enhance the sharpness of edges and denoise the neutron radiographs. Finally, the result of the adaptive fusion was embedded into a deblurring model based on total variation to further preserve the edges and restore more details during further restoration. In the experiments, it was demonstrated that our correction algorithm effectively preserved the prominent edges and the details in both the simulated and real neutron radiographs of thin and thick objects, exhibiting fewer ringing effects than traditional algorithms. It is worth mentioning that our work is primarily applicable and significant to neutron generators, but not for reactors or spallation neutron sources that can deliver much better images.

Measurement of cross-section of the 6Li(d,α)4He, 6Li(d,p)7Li, 6Li(d,p)7Li*, 7Li(d,α)5He, and 7Li(d,nα)4He reactions at the deuteron energies from 0.3 MeV to 2.2 MeV

The interaction of a deuteron beam with lithium is characterized by a high yield of neutrons, high energy of neutrons and a large variety of reactions. The reliable data of the reactions cross-sections are important for many applications, including radiation testing of advanced materials and equipment. The experimental data on the cross-sections differ greatly from one author to another; for a number of reactions there is no data on the cross-section in the databases. Measurements of the reactions cross-sections were carried out at the accelerator-based neutron source VITA at Budker Institute of Nuclear Physics (Novosibirsk, Russia) using an α-spectrometer. The 6Li(d,α)4He, 6Li(d,p)7Li, 6Li(d,p)7Li*, 7Li(d,α)5He, and 7Li(d,nα)4He reactions cross-sections at deuteron energies from 0.3 MeV to 2.2 MeV have been measured. The obtained data are presented tabular form.

NEBOAS: A Neutron yiElds Based On AcceleratorS application

The web-based application NEBOAS was developed to calculate the neutron yield of the accelerator-based neutron source MONNET (MONo-energetic NEutron Tower) at JRC-Geel. Neutrons are produced with p and d-induced reactions on particular targets. NEBOAS provides double differential neutron yields, as a function of the angle of neutron emission and energy. It provides also integrated neutron yields, and neutron energy spectra. Any projectile energy may be utilized, as long as it is covered by the nuclear data available on thin targets (that just degrade the projectiles' energy) or thick targets (that stop the projectiles). The calculation employs reaction cross-section evaluations from Evaluated Nuclear Data File (ENDF) or compilations of experimental measurements from the Experimental Nuclear Reaction Data (EXFOR), and stopping powers as recommended in the National Institute of Standards and Technology (NIST) or the Stopping and Range of Ions in Matter (SRIM-2013) databases.

Global characterization of a laser-generated neutron source

Laser-driven neutron sources are routinely produced by the interaction of laser-accelerated protons with a converter. They present complementary characteristics to those of conventional accelerator-based neutron sources (e.g. short pulse durations, enabling novel applications like radiography). We present here results from an experiment aimed at performing a global characterization of the neutrons produced using the Titan laser at the Jupiter Laser Facility (Livermore, USA), where protons were accelerated from 23 $\mathrm {\mu }$ m thick plastic targets and directed onto a LiF converter to produce neutrons. For this purpose, several diagnostics were used to measure these neutron emissions, such as CR-39, activation foils, time-of-flight detectors and direct measurement of $^7{\rm Be}$ residual activity in the LiF converters. The use of these different, independently operating diagnostics enables comparison of the various measurements performed to provide a robust characterization. These measurements led to a neutron yield of $2.0\times 10^{9}$ neutrons per shot with a modest angular dependence, close to that simulated.

Neutron emission characterization of IPR 14 MeV neutron generator

Accelerator-based neutron sources are being developed worldwide due to their wide applications. Institute for Plasma Research in India has developed an accelerator-based high-yield 14 MeV neutron generator based on deuterium-tritium (D-T) reaction to perform fusion-related benchmark experiments. This neutron generator is designed to produce 1012 neutrons per second in both mode continuous and pulse. The emission of neutrons from an accelerator-based neutron generator is not isotropic due to target geometry and angular dependency of reaction cross-section. A detailed and accurate description of neutron source emission and distribution is required to use such a neutron generator for various experiments. The neutron generator is characterized using experiments and simulations to accomplish the request. Energy and angular distributions of emitted neutrons are measured using a calibrated single crystal diamond (SCD) detector. Experimental data of neutron distributions are used to validate source neutron definition for Monte Carlo N Particle (MCNP) simulation. An accurate model of the Titanium Tritide (TiT) target and its holder geometry is prepared for the simulation. Finally, to validate the simulation model, simulated neutron fluxes are compared with measured ones by the SCD detector. This paper discusses the methodology adopted to characterize emitted neutrons and its results.

Influence mechanism of lithium surface oxidation on neutron yield of lithium target for accelerator-based neutron source

Lithium is generally adopted as the target to generate epithermal neutrons based on 7Li(p,n)7Be nuclear reaction for the application of accelerator-based boron neutron capture therapy (AB-BNCT). The stability of neutron yields and neutron energy spectrum are key factors for the therapeutic effects of the AB-BNCT. Owing to the active chemical properties of lithium, the surface oxidation formation of lithium may influence the soundness of the target system and the stability of neutron yields. Experimental and simulation methods were performed for a better understanding of the passivation layer formation after air exposure and the influence of the passivation layer on the neutron yield and energy spectrum. On one hand, X-ray photoelectron spectroscopy was adopted to study the chemical states of lithium after air exposure. On the other hand, particle and heavy ion transport code system (PHITS) was used to evaluate the effect of surface chemical states variation on neutron yield and energy spectrum. The simulation results indicate that the formation of Li2O with a thickness of 2 μm could mainly influence the neutron yields at the neutron emission angle of 20–90° with a maximum reduction of ∼ 10.4%.

Application of stoichiometric CT number calibration method for dose calculation of tissue heterogeneous volumes in boron neutron capture therapy.

Monte Carlo simulation code is commonly used for the dose calculation of boron neutron capture therapy. In the past, dose calculation was performed assuming a homogeneous mass density and elemental composition inside the tissue, regardless of the patient's age or sex. Studies have shown that the mass density varies with patient to patient, particularly for those that have undergone surgery or radiotherapy. A method to convert computed tomography numbers into mass density and elemental weights of tissues has been developed and applied in the dose calculation process using Monte Carlo codes. A recent study has shown the variation in the computed tomography number between different scanners for low- and high-density materials. The aim of this study is to investigate the effect of the elemental composition inside each calculation voxel on the dose calculation and the application of the stoichiometric CT number calibration method for boron neutron capture therapy planning. Monte Carlo simulation package Particle and Heavy Ion Transport code System was used for the dose calculation. Firstly, a homogeneous cubic phantom with the material set to ICRU soft tissue (four component), muscle, fat, and brain was modelled and the NeuCure BNCT system accelerator-based neutron source was used. The central axis depth dose distribution was simulated and compared between the four materials. Secondly, a treatment plan of the brain and the head and neck region was simulated using a dummy patient dataset. Three models were generated; (1) a model where only the fundamental materials were considered (simple model), a model where each voxel was assigned a mass density and elemental weight using (2) the Nakao20 model, and (3) the Schneider00 model. The irradiation conditions were kept the same between the different models (irradiation time and irradiation field size) and the near maximum (D1%) and mean dose to the organs at risk were calculated and compared. A maximum percentage difference of approximately 5% was observed between the different materials for the homogeneous phantom. With the dummy patient plan, a large dose difference in the bone (greater than 12%) and region near the low-density material (mucosal membrane, 7%-11%) was found between the different models. A stoichiometric CT number calibration method using the newly developed Nakao20 model was applied to BNCT dose calculation. The results indicate the importance of calibrating the CT number to elemental composition for each individual CT scanner for the purpose of BNCT dose calculation along with the consideration of heterogeneity of the material composition inside the defined region of interest.

Preparation and properties of chromium protective coatings on lithium targets for accelerator-based neutron sources

Chromium (Cr) with its high antioxidative properties acts as a protective layer for lithium targets in accelerator-based boron neutron capture therapy (AB-BNCT). Employing a controlled variable method, this study explores the impact of magnetron sputtering parameters (sputtering power, argon flow, and sputtering pressure) on the surface morphology and crystalline structure of Cr films. Under the optimal preparation parameters (300W, 20ssm, 0.5Pa), a dense and uniform film with a (110) phase was formed, and a compact Cr2O3 layer developed during air exposure. A 1 μm Cr coating provided effective protection for approximately 2 h, and increased thickness further improved the protection. Monte Carlo simulations revealed that a 1 μm Cr coating caused around 1.83% energy loss and 5.54% neutron yield reduction for 2.8 MeV protons, meeting AB-BNCT requirements with minimal impact on neutron angular distribution. This study demonstrates that by investigating the effects of magnetron sputtering preparation parameters on the surface morphology and crystalline structure of Cr coatings, optimal preparation parameters can be selected. The fabrication of a uniform and dense 1 μm Cr protective layer on a lithium target effectively prevents the degradation of the lithium target during short-term exposure to air, thereby enhancing its transportability and ease of handling.

A study on the treatment of brain tumors with BNCT using several moderators with different thicknesses

Boron neutron capture therapy (BNCT) is an effective binary radiation therapy that depends on nuclear capture reactions. In recent years, BNCT can be performed without a reactor owing to the development of accelerator-based neutron sources. A new BNCT irradiation facility is proposed, which is based on a 15 mA 2.5 MeV proton accelerator with a 100 μm thickness natural lithium target as a neutron converter. A great quantity of studies has shown that neutron beams with different spectra have unique therapeutic effects on tumors. An appropriate neutron beam for BNCT is obtained by Beam Shaping Assembly (BSA) and the moderator plays a main role in determining the BSA outlet beam spectrum. To figure out the dose distribution in phantom with various kinds of neutron spectrum modes during BNCT, a series of cases are calculated by MCNPX code. The results give a database for treatment of brain tumors with BNCT by using different moderators.

Study of accelerator-based epithermal neutron flux depending on Li and Be targets

The Li(p,n) and Be(p,n) reactions are widely used as accelerator-based neutron sources for boron neutron capture therapy (BNCT). Because the energy of the neutrons produced by these reactions is high for BNCT, a moderator is required to reduce the energy to the epithermal energy region. The neutron yields and energy spectra derived from the Li(p,n) and Be(p,n) reactions vary with the incident proton energy; a high proton energy results in a high neutron yield. However, a high proton energy can also increase the neutron energy, and in this case, a high neutron energy requires a thick moderator, which decreases the epithermal neutron flux. Notably, these tradeoffs are not completely clear. Furthermore, the upper limit of the epithermal neutron energy used in different BCNT facilities is not the same. Therefore, epithermal neutron flux estimation with applying an appropriate moderator is necessary for comparing epithernal neutron flux depending on the proton energy, target type, and epithermal neutron range definition.In this study, GEANT4 simulations were performed to determine effective moderator materials and thicknesses. After applying a moderator, we examined the epithermal neutron flux by changing the conditions that can influence the epithermal neutron flux, such as the proton energy, target species (Li and Be), and different definitions of the epithermal neutron region. The results showed that the epithermal neutron flux increases with the increasing proton energy until a certain level for both Li and Be targets. For proton energies above 3 MeV for the Li target and above 7 MeV for the Be target, the epithermal neutron flux decreases or its increase rate gradually decreases. And, when the upper energy limit of the epithermal neutron range is changed from 10 keV to 20 keV or 40 keV, the epithermal neutron flux is more than 109 cm−2 s−1 at a proton energy of 3 MeV for Li target and 8 MeV for Be target.

Accelerator-based neutron sources for BNCT

Accelerator-based neutron sources for BNCT

Estimation of nuclear properties in a deuteron–lithium target using JENDL/DEU-2020 for IFMIF and similar irradiation facilities

An accelerator-based neutron source powered by deuteron–Li (d–Li) reactions is among the most promising neutron sources for fusion materials irradiation facilities, such as IFMIF, IFMIF-DONES, and A-FNS. The yield estimation of neutrons and other particles is critical to the design of these facilities. Presently, the McDelicious code is employed for the aforementioned yield estimations, although it is not an open code. Recently, a deuteron-induced reaction library, JENDL/DEU-2020, was released by the Japan Atomic Energy Agency. Thus, employing benchmark tests, such as the Particle and Heavy Ion Transport code System (PHITS) + JENDL/DEU-2020, Monte Carlo N-Particle + JENDL/DEU-2020, and McDelicious, we confirm that PHITS (or Monte Carlo N-Particle (MCNP)) + JENDL/DEU-2020 effectively reproduces the experimental data and McDelicious results. We apply PHITS + JENDL/DEU-2020 to the yield estimations of the neutron, gamma, and charged particles in a Li target and to the yield estimation of the charged particles in the target backplate for A-FNS. The estimated production rates of tritium and 7Be during the continuous operation using a deuteron beam of 40 MeV and 125 mA are 1.4 × 1016 atoms/s (2.3 g/FPY) and 5.9 × 1014 atoms/s (0.21 g/FPY), respectively, and the estimated production rates of He and H in the target backplate for A-FNS are 3.2 × 102 and 1.8 × 103 appm/FPY, respectively.

Cavity tuning and beam commissioning of the radio-frequency quadrupole linac for a transportable neutron source

A transportable, compact, accelerator-based neutron source is under development at Xi’an Jiaotong University. At the heart of the linear accelerator is a radio-frequency quadrupole (RFQ) unit that has been designed to accelerate a proton beam from 30 keV to 2.5 MeV with a peak current of 12 mA and a duty factor of 3%. Recently, the manufacturing of the RFQ cavity has been completed. Cavity tuning, including field unevenness, resonant frequency, and rf coupling coefficient, was carried out after finishing brazing and assembly. The quadrupole perturbative component was controlled within ±2.0%, and the two dipole components were about ±2.5%. The value of the intrinsic Q was 9609, 89% of that of the theoretical simulation. We have completed the installation of the RFQ experimental setup and performed beam commissioning experiments. The output beam current reached 10.6 mA with a transmission efficiency of 93.8%.

Review of Cockcroft-Walton High Voltage Low Current DC Power Supplies

High-voltage power supplies are key for developing particle accelerators, neutron generators, x-ray systems, ion implantation, etc. This paper reviews various developed voltage low current power supplies based on Cockcroft-Walton voltage multiplier circuits for accelerator-based neutron sources and other industrial and medical applications. The paper presents a detailed review of the topology of the power supply used for the high voltage generation, the key design parameters, key active and passive components used in the High Voltage (HV) system, key experimental results, protection and measurement circuit used in the power supply and about the system performance. The improvement that occurred in the size and performance of the power supplies due to the development in the technology of the magnetic components, semiconductors and energy storage devices in successive years is apparent in the discussion. Finally challenges and possible solutions for the compact design and system performance are concluded for the development of high voltage low current power supplies.