Deuterium-deuterium Neutron Generator Research Articles

Progress in commissioning a neutron/X-ray radiography and tomography systems at IAEA NSIL

The Nuclear Science and Instrumentation Laboratory (NSIL) is currently establishing a Neutron Science Facility (NSF) based on two compact neutron generators: Deuterium-Deuterium (DD), resulting in 2.45 MeV neutrons, and Deuterium-Tritium (DT), resulting in 14.1 MeV neutrons, with maximum source intensities up to 5 × 106 n/s and 4 × 108 n/s over 4π, respectively. Neutron/X-ray radiography and tomography are two of the applications the NSF will be equipped with. In this paper, we report the state of the radiography and tomography system, and the neutron/X-ray radiography and tomography experiments we performed. Good radiographs and tomographs are obtained with X-ray. The spatial resolution of the X-ray radiography system is measured to be about 0.1 mm. The thermal neutron radiographs and fast neutron radiographs have proved that the DD neutron generator will be beneficial to demonstrate neutron radiography capabilities for educational purposes. Additionally, the detail of the collimator optimization of the DD neutron radiography was presented.

Measurement of the ionization yield of neutron-induced proton recoils in Tetramethylsilane

We report on a low energy measurement of the ionization yield in a Tetramethylsilane Time Projection Chamber (TPC) using 2.8 MeV neutrons from a deuterium-deuterium neutron generator. The proton recoil charge yield is measured at four different electric fields, finding a dependence that can be described by the Thomas-Imel model. By comparing the proton recoil yield to that obtained from γ-ray calibrations, a quenching factor is obtained for each electric field. These results demonstrate the feasibility of using room temperature organic ionisation detectors to detect MeV-scale neutrons in the proton-recoil channel.

Design and optimization of thermal neutron device based on deuterium-deuterium neutron generator

A thermal neutron device based on a portable D-D neutron generator has been designed and optimized by Monte Carlo method. The goal of optimal design is to maximize the thermal neutron flux at the output surface and improve the utilization efficiency of the neutron generator. The simulation of the reference device is experimentally verified. The contributions of different types and sizes of moderators, reflectors, and collimators to the thermal neutron intensity were simulated to obtain the optimum dimensions of the proposed design. Two parameters P and K were defined to describe the performance of device. It was determined that the final device uses high-density polyethylene (HDPE) as the moderator, nickel as the reflector, and HDPE as the collimator. The parameter K of the final device was 1.72. The P-value was 8 times that of the reference device. The thermal, epithermal + intermediate, fast, and total neutron flux on the output surface were 4.65, 2.52, 1.86, and 2.72 times of the reference device, respectively. The ratio of thermal to fast neutron flux ϕth/ϕfast increased from 0.52 to 1.30. The neutron irradiation device is supposed to be employed principally for neutron activation analysis, testing and calibration of neutron dosimeter and detector.

A feasibility study of a deuterium-deuterium neutron generator-based boron neutron capture therapy system for treatment of brain tumors.

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

SU‐G‐IeP4‐04: DD‐Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

Purpose:To optimize collimation and shielding for a deuterium‐deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT).Methods:Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm3 polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source‐side and target‐side openings) and aperture placements (13–39 cm from source‐side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system.Results:Analysis of primary beam indicated the beam‐width is linear to the aperture size. Increasing source‐side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings.Conclusion:The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well‐defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan time should be considered to determine the optimal design for specific NSECT applications.

SU‐E‐T‐21: A D‐D Based Neutron Generator System for Boron Neutron Capture Therapy: A Feasibility Study

Purpose:To investigate the feasibility of a deuterium‐deuterium (DD) neutron generator for application in boron neutron capture therapy (BNCT) of brain cancerMethods:MCNP simulations were performed using a head phantom and a monoenergetic neutron source, which resembles the point source in a DD generator that emits 2.45‐MeV neutrons. Source energies ranging from 5eV to 2.45MeV were simulated to determine the optimal treatment energy. The phantom consisted of soft tissue, brain tissue, skull, skin layer, and a brain tumor of 5 cm in diameter. Tumor depth was varied from 5–10 cm. Boron‐10 concentrations of 10 ppm, 15 ppm, and 30 ppm were used in the soft/brain tissues, skin, and tumor, respectively. The neutron flux required to deliver 60 Gy to the tumor as well as the normal tissue doses were determined.Results:Beam energies between 5eV and 10keV obtained doses with the highest dose ratios (3.3–25.9) between the tumor and the brain at various depths. The dose ratio with 2.45‐MeV neutrons ranged from 0.8–6.6. To achieve the desired tumor dose in 40 minutes, the required neutron flux for a DD generator was between 8.8E10 and 5.2E11 n/s and the resulting brain dose was between 2.3 and 18 Gy, depending on the tumor depth. The skin and soft tissue doses were within acceptable tolerances. The boron‐neutron interaction accounted for 54–58% of the total dose.Conclusion:This study shows that the DD neutron generator can be a feasible neutron source for BNCT. The required neutron flux for treatment is achievable with the current DD neutron technology. With a well‐designed beam shaping assembly and treatment geometry, the neutron flux can be further improved and a 60‐Gy prescription can be accurately delivered to the target while maintaining tolerable normal tissue doses. Further experimental studies will be developed and conducted to validate the simulation results.

A compact DD neutron generator–based NAA system to quantify manganese (Mn) in bone in vivo

A deuterium-deuterium (DD) neutron generator–based neutron activation analysis (NAA) system has been developed to quantify metals, including manganese (Mn), in bone in vivo. A DD neutron generator with a flux of up to 3*109 neutrons s−1 was set up in our lab for this purpose. Optimized settings, including moderator, reflector, and shielding material and thickness, were selected based on Monte Carlo (MC) simulations conducted in our previous work. Hand phantoms doped with different Mn concentrations were irradiated using the optimized DD neutron generator irradiation system. The Mn characteristic γ-rays were collected by an HPGe detector system with 100% relative efficiency. The calibration line of the Mn/calcium (Ca) count ratio versus bone Mn concentration was obtained (R2 = 0.99) using the hand phantoms. The detection limit (DL) was calculated to be about 1.05 μg g−1 dry bone (ppm) with an equivalent dose of 85.4 mSv to the hand. The DL can be reduced to 0.74 ppm by using two 100% HPGe detectors. The whole body effective dose delivered to the irradiated subject was calculated to be about 17 μSv. Given the average normal bone Mn concentration of 1 ppm in the general population, this system is promising for in vivo bone Mn quantification in humans.

MCNP modeling of a neutron generator and its shielding at Missouri University of Science and Technology

The shielding of a neutron generator producing fast neutrons should be sufficient to limit the dose rates to the prescribed values. A deuterium-deuterium neutron generator has been installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). The generator produces fast neutrons with an approximate energy of 2.5MeV. The generator is currently shielded with different materials like lead, high-density polyethylene, and borated polyethylene. An MCNP transport simulation has been performed to estimate the dose rates at various places in and around the facility. The simulations incorporated the geometric and composition information of these shielding materials to determine neutron and photon dose rates at three central planes passing through the neutron source. Neutron and photon dose rate contour plots at these planes were provided using a MATLAB program. Furthermore, the maximum dose rates in the vicinity of the facility were used to estimate the annual limit for the generator׳s hours of operation. A successful operation of this generator will provide a convenient neutron source for basic and applied research at the Nuclear Engineering Department of Missouri S&T.

Development of a transportable neutron activation analysis system to quantify manganese in bone in vivo: feasibility and methodology

This study was conducted to investigate the methodology and feasibility of developing a transportable neutron activation analysis (NAA) system to quantify manganese (Mn) in bone using a portable deuterium–deuterium (DD) neutron generator as the neutron source. Since a DD neutron generator was not available in our laboratory, a deuterium–tritium (DT) neutron generator was used to obtain experimental data and validate the results from Monte Carlo (MC) simulations. After validation, MC simulations using a DD generator as the neutron source were then conducted. Different types of moderators and reflectors were simulated, and the optimal thicknesses for the moderator and reflector were determined. To estimate the detection limit (DL) of the system, and to observe the interference of the magnesium (Mg) γ line at 844 keV to the Mn γ line at 847 keV, three hand phantoms with Mn concentrations of 30 parts per million (ppm), 150 ppm, and 500 ppm were made and irradiated by the DT generator system. The Mn signals in these phantoms were then measured using a 50% high-efficiency high-purity germanium (HPGe) detector. The DL was calculated to be about 4.4 ppm for the chosen irradiation, decay, and measurement time. This was calculated to be equivalent to a DL of about 3.3 ppm for the DD generator system. To achieve this DL with one 50% high-efficiency HPGe detector, the dose to the hand was simulated to be about 37 mSv, with the total body equivalent dose being about 23µSv. In conclusion, it is feasible to develop a transportable NAA system to quantify Mn in bone in vivo with an acceptable radiation exposure to the subject.