Accelerator-based Neutron Source Research Articles

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.

Development of a high current electron beam facility for compact accelerator neutron source target thermal tests

Target thermal capabilities are one of the major key points for CANS (Compact Accelerator-based Neutron Source) developments. Indeed, high energy (few MeV) and high average current (few tens of mA) proton or deuteron beams are required to produce high thermal neutron flux of 101³ n.cm⁻2.s⁻1 or greater. The targets must therefore absorb and dissipate a high thermal power. In order to make progress on their designs without deuteron or proton beams, we developed a compact electron beamline based on an Electron Cyclotron Resonance (ECR) source that permits testing of targets under relevant thermal conditions at power densities up 3 kW. cm⁻2.

Neutron flux evaluation algorithm with a combination of Monte Carlo and removal-diffusion calculation methods for boron neutron capture therapy.

In Japan, the clinical treatment of boron neutron capture therapy (BNCT) has been applied to unresectable, locally advanced, and recurrent head and neck carcinomas using an accelerator-based neutron source since June of 2020. Considering the increase in the number of patients receiving BNCT, efficiency of the treatment planning procedure is becoming increasingly important. Therefore, novel and rapid dose calculation algorithms must be developed. We developed a novel algorithm for calculating neutron flux, which comprises of a combination of a Monte Carlo (MC) method and a method based on the removal-diffusion (RD) theory (RD calculation method) for the purpose of dose calculation of BNCT. We present the details of our novel algorithm and the verification results of the calculation accuracy based on the MC calculation result. In this study, the "MC-RD" calculation method was developed, wherein the RD calculation method was used to calculate the thermalization process of neutrons and the MC method was used to calculate the moderation process. The RD parameters were determined by MC calculations in advance. The MC-RD calculation accuracy was verified by comparing the results of the MC-RD and MC calculations with respect to the neutron flux distributions in each of the cubic and head phantoms filled with water. Comparing the MC-RD calculation results with those of MC calculations, it was found that the MC-RD calculation accurately reproduced the thermal neutron flux distribution inside the phantom, with the exception of the region near the surface of the phantom. The MC-RD calculation method is useful for the evaluation of the neutron flux distribution for the purpose of BNCT dose calculation, except for the region near the surface.

Beam Dynamics Studies for the Target Beamlines of the High Brilliance Neutron Source

The High Brilliance Neutron Source (HBS) belongs to the class of High Current Accelerator based Neutron Sources (Hi-CANS). The driver Linac for the HBS must be able to reliably accelerate a 100 mA proton beam to an energy of 70 MeV [2, 1, 3]. The beam is sent pulse-by-pulse to three different targets via a multiplexer in the High Energy Beam Transfer (HEBT) section. Each individual proton pulse behind the multiplexer has a specific time structure in order to optimally serve the different instruments grouped around one target station. The main development steps of the HEBT are the development of a three-field septum magnet, which is an essential part of the multiplexer magnet system, the beam dynamics integration of the multiplexer magnet system into the beamline, and the ion-optical layout of the individual target beamlines.

Design and experiments of a compact electrostatic low energy beam transport with a double Einzel-lens for a transportable neutron source.

A transportable, compact, accelerator-based neutron source is under development at Xian Jiaotong University. An electrostatic low energy beam transport (LEBT) structure with a double Einzel-lens setup was adopted due to its short length and low power consumption. It can transport a pulsed proton beam to the radio frequency quadrupole with a required beam current of 15 mA and an energy of 30keV. We performed detailed structure optimization and beam tracking to achieve beam matching and small emittance growth. In addition, the fast chopper, beam steering, and diagnostic devices are integrated into the LEBT. The fabrication and assembly of the proton injector have been completed, and beam commissioning was carried out to measure the beam current and Twiss parameters. The design strategy, beam simulation, and experimental results are presented and discussed in this paper.

Development of an accelerator-based neutron source to prototype Mo-99 production, Part II: A liquid LBE loop under a high vacuum

To provide US domestic supply of Mo-99 without using high-enriched uranium (HEU), a subcritical uranium target assembly (UTA) is irradiated by an accelerator-based neutron source to create Mo-99 through fission. Part I of this work discusses the design of a liquid lead–bismuth eutectic (LBE) windowless target for an accelerator-based neutron source development. Part II discusses how to couple this windowless target to an accelerator operating at an ultra-high vacuum and the subcritical UTA cooled by water at room temperature.Due to the windowless design of the target, the liquid LBE flow shares an ultra-high vacuum (<1.3 × 10−7 Pa or 10−9 Torr) space with the accelerator. As a result of this shared vacuum space, the LBE system must operate at a high vacuum (10−3 ∼ 10−6 Pa or 10−5 ∼ 10−8 Torr). A magnetic rotary motion feedthrough unit utilizes magnetic fluid to allow rotation of the pump while maintaining a high vacuum environment. Prior to testing the LBE system under vacuum, a pump curve measurement is performed to estimate flowrate in the system. This measurement also generates data on orifice loss coefficients, which are compared to correlations in literature. The second experiment investigates vacuum level in the LBE system during operation. High vacuum is maintained (10−3 ∼ 10−5 Pa or 10−5 ∼ 10−7 Torr) during system operation, and a residual gas analyzer (RGA) scan shows that partial pressures of residual gases in the LBE system lower over the duration of LBE system operations. The third experiment investigates the gravity driven liquid LBE flowing out of the target chamber in the return line, which is partially full. If the liquid LBE is not drained quickly enough, flooding in the target chamber could occur. The coefficient n in the Manning equation is found to be around 0.008 s/m1/3. The last experiment performed is a demonstration that a vacuum jacket could provide sufficient thermal insulation to allow coupling between 300 °C LBE loop and a water tank at room temperature. The results from these experiments have influenced the development of the neutron source for the future commercial scale Mo-99 production system.

Development of an accelerator-based neutron source to prototype Mo-99 production, part I: A liquid LBE windowless target

Molybdenum-99 (Mo-99)’s decay product, technetium-99 (Tc-99 m), is one of the most critical isotopes for medical diagnostics. To provide U.S. domestic supply of Mo-99 without using high-enriched uranium (HEU), a subcritical uranium target assembly (UTA) is irradiated by an accelerator-based neutron source to create Mo-99 through fission. This study discusses the development of the accelerator-based neutron source. The high-energy electrons from the accelerator irradiate a liquid lead-bismuth eutectic (LBE) target to produce neutrons. Part I of this work focuses on numerical and experimental analysis towards the development of a liquid LBE windowless target. Unlike the existing windowless targets in literature, the current design creates a vertical free surface for a beam to irradiate. First, a hydrodynamic analysis of the LBE windowless target is performed. Simplified analytical calculations are assisted by 2D computational fluid dynamics (CFD) simulations to design the target, with the focus on eliminating recirculation zones and avoiding cavitation. With the optimized geometry, the experimental study is performed to investigate the flow hydrodynamics using liquid LBE. The experiments (1) compare pressure drop in the system to correlation predictions; (2) visualize the free surface liquid LBE flow from the beam view; (3) validate the LBE flow profile using temperature sensitive paint from the side view; and (4) validate the liquid LBE film thickness using gamma densitometer measurements. Second, the power handling capability of the designed windowless target is investigated. The divider plate in the current design is susceptible to overheating due to the thin LBE film in front. As LBE erosion and corrosion is likely to occur at an LBE velocity of 2.0 m/s and temperature above 500 °C, a power limit of 10 kW of beam power was established to prevent this corrosion from occurring, which is calculated by a Nusselt number correlation. The divider plate surface temperature at 10 kW agrees well with the 3D CFD simulation results. Part I demonstrates the fundamental physics in liquid LBE windowless target design and associated testing. A companion paper, Part II will demonstrate how to couple this windowless target into the Mo-99 production system, including an accelerator system operating under an ultra-high vacuum and the UTA cooled by water at room temperature.

Plasma window performance and scaling for an accelerator-based neutron source.

A "plasma window" was developed for use with a SHINE accelerator-based neutron source. In this work, the design of the plasma window is presented along with results demonstrating successful operation in deuterium over a range of aperture diameters (5-12mm), gas flow rates (3.5 to 12.5 slm), and input powers (10-55 kW). An empirically determined relationship between the aperture diameter, deuterium gas flow rate, input power, and plasma window pressure differential is also presented.

Comparison of biological weighting functions to estimate the microdosimetric RBE in BNCT.

In the framework of the MUNES project, a prototype accelerator-based thermal neutron source was developed and installed at the Legnaro National Laboratories of INFN. The microdosimetric characterization of this radiation field was performed with a Tissue-Equivalent Proportional Counter with interchangeable cathode walls, either doped with 100ppm of 10B or without boron doping. A suitable subtraction procedure allowed to discriminate the gamma, neutron and BNC dose components (Selva etal., 2022, Appl. Radiat. Isot. 182, 110144). The measured microdosimetric spectra can be weighted with a biological weighting function to estimate the Relative Biological Effectiveness of the radiation field, for the purpose of intercomparison between different thermal neutron sources. This work compares, therefore, the biological doses resulting from four different weighting functions applied to the same initial microdosimetric spectrum, discussing strengths and limitations of each of them.

Current development status of accelerator-based neutron source for boron neutron capture therapy

Recently, progress in technology for accelerator-based neutron sources has increased attention regarding boron neutron capture therapy (BNCT). BNCT is a type of radiotherapy that combines neutrons and boron drugs and is expected to be used in the treatment of refractory and recurrent cancers. Owing to the need for high-intensity neutrons in treatment, compact accelerator-based neutron sources applicable to BNCT are being developed worldwide. These current projects utilize cyclotrons, linear accelerators, and electrostatic accelerators as accelerators for BNCT devices. Beryllium and lithium are the main target materials for neutron generation. The accelerators for BNCT device are required to accelerate charged particles with an average current ranging from a few milliamperes to a few tens of milliamperes in order to generate neutrons of sufficient intensity for the treatment. Moreover, the target systems require technologies and mechanisms that can withstand the large heat load produced by high-power beam irradiation and prevent blistering. This review outlines and explains the accelerator neutron sources for BNCT and the requirements for the components of each device, such as the accelerator, target material, and beam shaping assembly. In addition, various development projects for accelerator-based BNCT devices worldwide are introduced.

Dosimetry for boron neutron capture therapy developed and verified at the accelerator based neutron source VITA

The method of boron neutron capture therapy for malignant tumors, proposed in 1936, is beginning to enter clinical practice. The development of dosimetry tools for characterization of therapeutic mixed neutron-photon beam and assessing the patient’s response to treatment is becoming relevant. In this work, a number of dosimetric techniques have been developed: a compact neutron detector with a pair of cast scintillators, one of which is enriched with boron, to measure the boron dose and the γ-ray dose; cell dosimeter for measuring the sum of the equivalent dose of fast neutrons and the equivalent nitrogen dose; prompt γ-ray spectroscopy for in situ measurement of boron dose in real time; epithermal neutron flux monitor for measuring the epithermal neutron flux. Their verification carried out on the accelerator based neutron source VITA showed that they can become convenient and reliable tools for characterization of neutron beam and assessing the patient’s response to treatment.