Monoenergetic Sources Research Articles

Device response principles and the impact on energy resolution of epitaxial quantum dot scintillators with monolithic photodetector integration

Epitaxial quantum dot (QD) scintillator crystals with picosecond-scale timing and high light yield have been created for medical imaging, high energy physics and national security applications. Monolithic photodetector (PD) integration enables the sensing of photons generated within the waveguiding crystal and allows a wide range of scintillator-photodetector coupling geometries. Until recently, these doubly novel devices have suffered from complex, high variance responses to monoenergetic sources which significantly reduces their precision and accuracy. The principles governing the overall device response have now been discerned and embodied by an expression derived within a geometrical optics framework which considers optical properties, surface roughness and photodetector coupling geometry. Response variation due to these factors was sufficiently reduced to obtain material-related energy resolution values of 2.4% with alpha particles. These findings place energy resolution alongside luminescence timescale, photon yield, and radiation hardness as outstanding properties of these engineered materials.

Determination of neutron spectrum based on artificial neural network using liquid scintillation detector EJ-301.

This paper focuses on the neutron spectrum measurement using a liquid scintillation detector, where the neutron spectrum could be identified and unfolded from the light output distribution of the EJ-301 liquid scintillation detector through a linear artificial neural network (ANN). The response functions of the EJ-301 detector for monoenergetic neutron sources, as well as the light outputs, have been simulated and calculated by Monte Carlo procedure FLUKA. The linear ANN was trained and tested through the simulated data, where response functions were set as the input of ANN and the corresponding neutron spectra were output. Therefore, the neutron spectrum-unfolding model was created. This spectrum-unfolding model was tested through the light outputs induced by monoenergetic neutrons and the random superposition of them. Unfolding results show that this model could identify the information of the neutron spectrum accurately from the light outputs of a liquid scintillation detector. Moreover, the EJ-301 detector was used to measure the radioactivity of 252Cf, and the pulse height distribution induced by neutrons was derived through the charge-comparison method to remove the influence of gamma rays. The measured pulse height distribution was unfolded by the trained model, and measured results show that the unfolded neutron spectrum of 252Cf was consistent with the reference one. This paper presents the feasibility that the unknown neutron spectrum could be identified and confirmed through a linear neural network trained by simulated monoenergetic neutron response functions, which could be a candidate of choice for the determination of the neutron spectrum.

Optimization and experimental validation of deuterium-tritium neutron monoenergetics for diamond detector testing

Diamond detectors are excellent tools for measuring deuterium-tritium (D-T) neutron energy spectra. It has great potential for application in the field of magnetic confinement fusion. However, the current experimental values of the energy resolution of diamond detectors are much lower than the theoretical estimates. The purpose of this study is to find a method for experimental optimization. For this purpose, the D-T neutron energy distribution of the accelerator is theoretically analyzed and simulated. A method to increase the deuteron energy to obtain a better monoenergetic neutron source is proposed. This strategy was validated experimentally. The energy resolution of diamond for D-T neutrons achieves 1.21%, a Full Width at Half Maximum of 93 keV at 7.65 MeV. It is the highest resolution recorded to date. These results demonstrate the feasibility of the optimization method.

Comparison of photon interaction coefficients for tumor compositions and healthy tissues simulated by Monte Carlo

This study compares interaction coefficients of photons (namely, mass attenuation, energy absorption and energy transfer coefficients) for some tumor tissues with simlar healthy tissues. Monte Carlo calculations were performed for adenoidcystic carcinoma, melanoma, rectal adenocarcinoma, sarcoma and squamous cell lung carcinoma. The simulation model involved a monoenergetic point source producing a pencil beam. Depending on the parameter under study, average flux, energy flux, or dose deposition from photons that travels in an absorber were scored in the range of 10 keV–20 MeV energy using MCNP6.1. The same model was used to compute the interaction coefficients of health tissues (namely, gastrointestinal tract-small intestine wall, lungs-parenchyma, salivary glands, skin, soft tissue (female+male)) for comparison purposes. The results showed that, depending on the contents of the tumor samples, photon interaction coefficients of a tumor may differ from those of a similar healthy tissue. This behavior was distinct, especially for energies up to 100 keV, and was attributed to the presence of relatively higher atomic number elements in the absorber.

Evaluating deposited radiation energy amount and collision quantities of small-molecule radiosensitizers through Monte Carlo simulations

This study investigates the photon interaction mechanism of various small molecule radiosensitizers, including Hydrogen Peroxide, Nimorazole, 5-Fluorouracil, NVX-108, and others, using the MCNP 6.3 Monte Carlo simulation code. The simulations focused on quantifying the linear attenuation coefficients, mean free path, and accumulation factors of these radiosensitizers, as well as their interactions in a simulated spherical water phantom irradiated with a 100 keV mono-energetic X-ray source. Our findings reveal significant variations in deposited energy, collision events, and mean free path among the radiosensitizers, indicating different efficacy levels in enhancing radiation therapy. Notably, NVX-108 demonstrated the highest energy deposition, suggesting its potential as a highly effective radiosensitizer. The study also examined the individual attenuation properties of these radiosensitizers against energetic photons, with NVX-108 showing the highest attenuation coefficient and a shorter mean free path, further supporting its superior potential in effective radiosensitization. It can be concluded that NVX-108 has higher interaction tendency with the energetic photons comparing other small-molecules under investigation.

Reconstruction of indoor gamma-ray background spectrum for HPGe detectors

Background reconstruction stands as a critical technique for diverse domains, such as signal isolation for gamma spectroscopy, pre-testing of radiation shielding designs, and environmental monitoring. Traditional methods aim to seek the optimal intensity of the background source under a fixed spatial distribution. However, their efficacy is often constrained due to the limited number of variable parameters, resulting in inadequate degrees of freedom and suboptimal performance. This paper introduces an innovative method for background reconstruction, aiming to overcome the limitations of existing approaches. The proposed technique involves the simultaneous adjustment of both intensities and shell thicknesses of a series of monoenergetic sources, modeled as spherical shells. The reconstructed spectrum is conceptualized as a superposition of simulated spectra from these sources, with the optimized parameters being the intensity and thickness of each source. The objective is to minimize the disparity in regions of interest (ROIs) between reconstructed and target spectra through a multivariate nonlinear optimization process, incorporating distinct weight factors for each ROI. The efficacy of the proposed method is validated using a simulation-generated theoretical spectrum. Results demonstrate successful reconstruction, with a high overall concordance between reconstructed and target spectra, showcasing reliability of the method. Subsequently, the method is applied to the reconstruction of an experimental spectrum, and a comparative analysis with a fixed spherical shell thickness method reveals the superior agreement of our proposed method with the experimental spectra.

Gamma-Ray Shielding Capacity of Bi2O3-SiO2-B2O3 Glass Powders with Different Bi2O3 Contents

In the present work, features of the ionizing radiation shielding of Bi2O3-SiO2-B2O3 glass powders with different Bi2O3 contents (45–60 mass%) has been investigated by using GAMOS (version 6.2). The validation of generated GAMOS simulation geometry has been provided by comparing the results with standard XCOM data for mass attenuation coefficients of glass powders. The Monte Carlo simulations were used to score photons that traveled in an absorber within the energy range of 0.01 MeV to 20 MeV, depending on the parameter under study. The simulation model involved a monoenergetic point source producing a pencil beam, absorber, and detector. We have calculated the mass attenuation coefficient (MAC), Half-value layer (HVL), Tenth-value layer (TVL), and Mean Free Path (MFP). The greatest linear attenuation coefficients in the whole energy range are related to Bi60 and the lowest were to Bi45. The obtained results were compared and these results are in good agreement with the obtained values from the XCOM program

Monte Carlo calculated photon interaction coefficients for several body tissues.

Absorption of energy in body tissues because of radiation interactions may induce harmful outcomes such as cancer and hereditary effects due to a variety of damages in the integrity and activity of the cells. This study presents Monte Carlo calculated $\boldsymbol{\mu} /\boldsymbol{\rho}$, ${\boldsymbol{\mu}}_{\boldsymbol{en}}/\boldsymbol{\rho}$ and ${\boldsymbol{\mu}}_{\boldsymbol{tr}}/\boldsymbol{\rho}$ values of some common tissues and organs found in the human body (namely, adipose tissue, blood, bone-cortical, brain-grey/white matter, breast tissue, eye lens, lung tissue, muscle-skeletal, ovary, soft tissue and testes) as well as water for comparison purposes. The simulation model involves a monoenergetic point source producing a pencil beam where, depending on the parameter under study, particle flux, energy flux or absorbed dose from photon interactions are scored in the range of 10keV to 20MeV energy. The simulations were performed using the Monte Carlo package MCNP6.1 and provided $\boldsymbol{\mu} /\boldsymbol{\rho}$, ${\boldsymbol{\mu}}_{\boldsymbol{en}}/\boldsymbol{\rho}$ and ${\boldsymbol{\mu}}_{\boldsymbol{tr}}/\boldsymbol{\rho}$ values. The data produced in this study were compared with theoretical photon attenuation data from the XMUDAT database and demonstrated good agreement. The results, which are based on a simple model geometry and pure elemental compositions, indicate that this approach can be applied to evaluate $\boldsymbol{\mu} /\boldsymbol{\rho}$, ${\boldsymbol{\mu}}_{\boldsymbol{en}}/\boldsymbol{\rho}$ and ${\boldsymbol{\mu}}_{\boldsymbol{tr}}/\boldsymbol{\rho}$ in a broad energy range for any element, compound or mixture.

DEVELOPMENT AND TESTING OF A COMPUTER MODEL OF A SYSTEM FOR FORMING NEUTRON FLUXES ON A LINEAR ELECTRON ACCELERATOR

A computer model of the system for forming neutron fluxes at the output of a linear electron accelerator has been developed in the Geant4 programming environment. The system for forming neutron fluxes consists of a monoenergetic neutron source, a graphite reflector, a polyethylene moderator, lead protection and detector. With the help of the model, a number of virtual experiments on 107 primary neutrons were carried out for the energy range of the neutron source from 0.1 to 1.1 MeV. The dependence of the ratio of the number of neutrons falling on the detector with a reflector to the number of neutrons without a reflector is determined, on the energy of the neutron source. Energy spectra of neutrons falling on the detector are determined. It was established that the graphite reflector not only increases the number of particles falling on the detector, but also moderated the energy spectrum of neutrons. The neutron background in the accelerator bunker when using a reflector decreases from 2 to 7 times, depending on the energy of the neutron source.

Measurement of cross section for the 232Th(n, f) reaction using a time projection chamber

Accurate cross sections of neutron induced fission reactions are needed in the development of nuclear science and technology. The time projection chamber (TPC) is viewed as the potential detector to improve the measurement accuracy of the fission cross sections based on its track reconstruction and particle identification capabilities. The TPC developed by the back-streaming white neutron source (Back-n) team of China Spallation Neutron Source (CSNS) was introduced to verify its ability in measuring the cross section of fission reaction. Using our TPC, and based on the mono-energetic neutron source at Peking University and 232Th/238U samples, we have measured the cross section of the 232Th(n, f) reaction at 5.0 MeV. The fission events have been well distinguished from the background alpha events. The present cross section is consistent with the evaluation data from different libraries, which indicates that the fission cross section measurement using our TPC is feasible. By reducing the distance between the two fission samples, the large uncertainty of the present result will be reduced in future experiments.

Development of DynamicMC for PHITS Monte Carlo package.

Previously, we have developed DynamicMC for modeling relative movement of Oak Ridge National Laboratory phantom in a radiation field for the Monte Carlo N-Particle package (Health Physics. 2023,124(4):301-309). Using this software, three-dimensional dose distributions in a phantom irradiated by a certain mono-energetic (Mono E) source can be deduced through its graphical user interface. In this study, we extended DynamicMC to be used in combination with the Particle and Heavy Ion Transport code System (PHITS) by providing it with a higher flexibility for dynamic movement for an anthropomorphic phantom. For this purpose, we implemented four new functions into the software, which are (1) to generate not only Mono E sources but also those having an energy spectrum of an arbitrary radioisotope (2) to calculate the absorbed doses for several radiologically important organs (3) to automatically average the calculated absorbed doses along the path of the phantom and (4) to generate user-defined slab shielding materials. The first and third items utilize the PHITS-specific modalities named radioisotope-source and sumtally functions, respectively. The computational cost and complexity can be dramatically reduced with these features. We anticipate that the present work and the developed open-source tools will be in the interest of nuclear radiation physics community for research and teaching purposes.

Kinetic Energy of Secondary Particles Produced from Various Electromagnetic Interactions in Water: A Monte Carlo Study

Advancements in imaging systems including Electronic Portal Imaging Devices (EPIDs) play a great role in radiotherapy treatment. It was developed as a verification tool for patient setup during radiotherapy sessions and also become a promising tool for the determination of the accurate placement of radiation beams. However, as part of quality assurance, individual patient treatments are often verified by patient-specific quality control measurements such as before treatment (pretreatment) or during treatment (in vivo). It has been shown that in vivo dosimetry using an electronic portal imaging device (EPID) is an effective QC tool to detect errors and this method has been clinically applied to various treatments. The introduction of advanced EPID technology has led to an interest in its application for dose conformation and dose deposition. Moreover, dose deposition is subject to uncertainties due to several factors, including the presence of secondary particles. Thus, knowing the physical processes that produced the secondary particles as well as their average kinetic energy will help to provide valuable information about the effective filtering of these particles or the possible use of these particles for other applications. In this study, Monte Carlo simulations are performed to determine the average kinetic energy of detected secondary particles, specifically photons, electrons, and positrons produced by each particular physical interaction as a function of detector position using GATE v9.0. The virtual radiotherapy set-up is composed of the box water phantom, which is the target in the simulations with a dimension of 20 cm × 20 cm × 20 cm, an EPID system (detector), and a beam source in which it uses three (3) beams situated at varying positions with an energy of 6 MeV. The monoenergetic pencil beam source is placed 90 cm away from the center of the target and is directed toward the target (+x-axis) while the EPID (detector) is set as 120 cm SDD (source-to-detector distance). Moreover, the photon beam with 10 million primaries is set with varying field sizes of 1 cm × 1 cm, 3 cm × 3 cm, 6 cm × 6 cm, and 9 cm × 9 cm. Overall, the results show that the highest average kinetic energy among secondary particles produced by each physical interaction are electrons coming from Compton scattering (∼ 3 MeV), followed by positrons and electrons from pair production (∼ 2.4 MeV), photons from annihilation and bremsstrahlung (∼ 0.5 MeV), and electrons from ionization (∼ 0.13 MeV).

Comparison of calibration coefficients for a vinten ionization chamber simulated using four Monte Carlo methods

The Vinten 671 ionization chamber (VIC) was modelled using Monte Carlo (MC) programs EGSnrc, Penelope, and TOPAS. Several national measurement institutes have VICs with well-characterized response relationships and have measured calibration coefficients for many radionuclides. Twelve radionuclides with various decay emissions were assessed as well as 14 monoenergetic photon sources and 10 monoenergetic electron sources. Calibration coefficients were calculated based on the energy deposited in the simulated VIC nitrogen gas volume and compared to experimental values from the literature.

Gamma-ray attenuation properties of polymer biomaterials: Experiment, XCOM and GAMOS results

This study aims to compute and measure the linear attenuation coefficients of two biomaterials used in the human body as synthetic materials. These are used as bone cement and have applications in contact lens production and dentistry (coded as P1) also glass ionomer cement is a frequently preferred material in restorative dentistry applications (coded as P2). In the computational part of the study, a mono-energetic photon source between 1 keV and 20 MeV was used in the simulations performed using the GAMOS 6.2 program. In the experimental part, the linear attenuation coefficients of polymer materials were measured by using different radioactive sources at 6 different certain energies ranging from 511 to 1332 keV. A gamma spectrometer consisting of a NaI (Tl) detector was used for this purpose. Using the XCOM code, the experimental and simulation results were compared to the theoretical results, and a very good agreement was found between them. The simulation results demonstrated that the Monte Carlo technique can be used as an alternative to calculating the mass attenuation coefficients at the desired gamma energy, particularly for physically difficult to produce samples.

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

M-TAG: A modular teaching-aid for Geant4

Geant4 is a versatile Monte Carlo radiation transport simulation toolkit with a steep learning curve. This work introduces a user-code called M-TAG (Modular Radiation Teaching-Aid for Geant4), built on top of Geant4. M-TAG is designed to help gradually introduce the Geant4 toolkit to new users.The goal of Geant4 is to record quantities from the simulated radiation as it is transported through geometries. M-TAG simplifies the inclusion of new geometric elements and detector components in the simulation by including new classes. M-TAG also provides basic validated examples for some common detector development tasks. Geant4 intercom modules, called messenger classes, manage these classes. To validate M-TAG, simulations were performed to calculate the range of positrons in water. One hundred million decays at the center of a water-filled sphere with a radius of 1 m were allowed for fluorine-18, carbon-11, oxygen-15 and gallium-68. These results were compared to literature values. An inexperienced Geant4 user was tasked with creating a simulation model for a plastic scintillator-based detector and conducting basic tests to assess the effectiveness of M-TAG as a teaching tool. The simulation involved calculating the dose to the detector's sensitive volume using a 2x2 cm planar monoenergetic photon source spanning energies from 20 to 100 keV. One billion particles were simulated twice: once with the actual detector geometry and once with the sensitive volume replaced by water. The validity of M-TAG was also verified by computing dose ratios and comparing them with mass-attenuation ratios obtained from NIST XCOM data sets.The mean positron travel distances were within the distribution of literature values. Simulated positron energy spectra means were within 1.8% of literature means. Simulated dose ratios agreed with literature values within uncertainties.We have developed and verified a modular Geant4 teaching aid called M-TAG. It was used to introduce a new user to Geant4, who used it to perform further validation simulations.

Neutron spectroscopy with TENIS using an artificial neural network

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

Air kerma rate from radionuclides distributed in forest ecosystem

This study evaluates the air kerma rate in radioactively contaminated forests. The air kerma rates created by plane sources of monoenergetic photons in the energy range 0.02–3 MeV located at different depths in soil up to 50 g cm−2 and at different heights in the forest medium from 0.05 to 50 m were calculated using numeric solution of the transport (Boltzmann) equation. To simplify the practical use of the results obtained by solving the Boltzmann equation, the study additionally includes approximation formulae for calculating air kerma rate separately from contaminated soil, crowns and trunks of trees in a forest ecosystem for 20 radionuclides - fission products that significantly contribute to the external dose. Biomaterial of trunks and crowns was modeled as uniformly distributed in corresponding layers and homogeneously mixed with air. Different distributions of radionuclides in soil were considered including plane source located at different depths, exponential and uniform distribution. Based on the results, the effect of forest biomass presence as an absorbing and scattering medium on the air kerma rate at 1 m above soil was evaluated. The estimated relative difference in air kerma rate at 1 m above soil in the forest medium and in free air for monoenergetic photon sources with energies 0.1 MeV, 0.66 MeV and 3 MeV did not correlate significantly with the energy of photons. Its maximum value in forest medium with biomass density of 5 kg m−3 was 15–20% for the source at soil depth ∼0.3 g cm−2, decreasing to less than 5% when it is at soil depth greater than 7 g cm−2. An example calculation of the air kerma rate dynamics is presented for the initial period after radioactive fallout considering weathering processes (rainfall and wind action) that contribute to the transfer of activity from the canopies to the forest floor. The differences in air kerma rate values, as an integral characteristic of the gamma radiation field from a radioactive cloud in the forest and in the open area, were evaluated.

True coincidence summing correction for a BEGe detector in close geometry measurements

The true coincidence summing correction factor for a Broad Energy Germanium detector has been calculated at far and close geometry set-up using radioactive γ-ray sources. The correction factors were calculated using both experimental and analytical methods. Geant4 simulation was done to calculate the full-energy peak and total efficiencies of the detector. Standard, as well as fabricated mono-energetic γ-ray sources, were used for the γ-ray efficiency measurements. The simulated efficiencies of mono-energetic γ-ray sources were matched to the experimental γ-ray efficiencies by optimizing the detector parameters. The same parameters were used to obtain the full-energy peak and total efficiencies for γ-rays of current interest. Analytical and experimental correction factors were found to agree well with each other. The coincidence summing effect is found to be significant for source-to-detector distances less than 5 cm.

Comparison of MCNPX and EGSnrc Monte Carlo Codes in the Calculation of Nano-Scaled Absorbed Doses and Secondary Electron Spectra around Clinically Relevant Nanoparticles

Purpose: Absorbed dose enhancement due to the presence of high atomic number Nanoparticles (NP)s has been estimated and modeled by Monte Carlo (MC) simulation methods. In the current study, two MC codes of Monte Carlo N‐Particle eXtended (MCNPX) and EGSnrc codes were compared by calculation of secondary electron energy spectra and nano-scaled dose values around four types of spherical NPs.
 Materials and Methods: The MC model was composed of a spherical nanoparticle with a diameter of 50 nm and mono-energetic sources of photons with energies of 30,60, and 100 keV. The secondary electrons emitted from the nanoparticle were scored on the nanoparticle surface and the delivered dose to water around the nanoparticle was tallied using concentric shells with a thickness of 25 nm. Four different elements were used as materials of NPs, including Gold, Bismuth, Gadolinium, and Hafnium.
 Results: Our results showed a considerable difference in the number of emitted electrons per incident photon between the two codes. There were also discrepancies between the two codes in the energy spectra of secondary electrons. Calculated radial dose values around NPs in nano-scale had a similar pattern for both codes. However, significant differences existed for some elements.
 Conclusion: It can be concluded that the results of nano-scaled MC modeling for nanoparticle-based radiation therapy are dependent on the code type and its algorithm for electron transport as well as exploited cross-section libraries.