Monte Carlo N-Particle Simulation Research Articles

Parametric optimization for estimating beta detection efficiency in thin plastic scintillation detector

A condenser air removal system plays a crucial role in radiation monitoring within nuclear power plants, specifically for detecting potential leaks of radioactive material from primary-to-secondary systems. This system assesses the leakage rate (LR) by measuring the emission of β particles from noble gases using a thin plastic scintillation (TPS) detector, which is specifically designed for counting purposes. Among the factors used in LR calculation, the detection efficiency (k) must be experimentally determined in advance. However, this is challenging because of the limitations posed by radiation safety regulations. Moreover, simulation-based calculations are difficult to achieve because only a portion of the β-ray energy is transferred in the TPS detectors, resulting in a lack of prior energy information. To overcome these challenges, this study employed a multi-objective parameter optimization technique to simultaneously calculate the parameters required for the β-ray detection response function and energy calibration equation. Subsequently, k values were calculated using the optimization results. The optimization performance was demonstrated by comparing the measured spectra of the TPS detector with the spectra obtained from the Monte Carlo N-particle simulation. Furthermore, we confirmed that the calculated k values agreed with the reference values.

A modular, high dynamic range passive neutron dosimeter and imaging diagnostic.

The multi-decade neutron dosimeter and imaging diagnostic (MDND) is a passive diagnostic that utilizes the polyethylene (n, p) nuclear reaction to enhance the diagnostic's sensitivity for time and energy integrated neutron measurements in the range of 2.45-14.1MeV. The MDND utilizes a combination of radiochromic film, phosphor image plates, and solid-state nuclear track detectors, with the goal of providing several orders of magnitude of dynamic range in terms of measured neutron fluence. The diagnostic design was guided by simulations in the Monte Carlo N-Particle (MCNP) transport code to determine the optimum thickness of the polyethylene convertor for maximum proton fluence incident on the detection medium as a function of incident neutron energy. In addition, the simulation results of complete diagnostic assemblies, or "stacks," were used to determine the total dynamic range of an MDND in terms of measured neutron source yield, which was found to be between around 107 and 1015 emitted into 4π with the detector located 1m away from the source. Complimentary to these simulations, individual detectors within a stack were simulated and analyzed to determine response as a function of neutron energy and yield. This work presents the diagnostic design, MCNP simulation results, and analysis of expected signals for varying neutron sources.

A high-intensity, low-energy heavy ion source for a neutron target proof-of-principle experiment at LANSCE

The capability to directly study neutron capture reactions on radionuclides with half-lives on the of order minutes would allow key cross section measurements in nuclear astrophysics and energy applications. However, such experiments with stationary targets are currently impossible because of signal detection and target sample fielding issues. To overcome these challenges, a neutron target facility is being developed to permit neutron capture experiments on unstable isotopes in inverse kinematics at the Los Alamos Neutron Science Center (LANSCE). This next-generation facility will consist of a heavily moderated, high-intensity spallation neutron target coupled with a radioactive ion beam storage ring. A proof-of-principle experiment is underway to demonstrate this neutron target concept at LANSCE in the near term with stable ions and without a storage ring. Here, mA-level beams of 10-50 keV heavy ions exhibiting large resonant neutron capture cross sections will be generated by a new ion source, transported through a large-volume neutron moderator surrounding an adjacent spallation target driven by the LANSCE accelerator, and collected downstream of the moderator for subsequent decay-counting measurements. The neutron density within the moderator will be obtained from these decay yields and compared with Monte Carlo N-Particle (MCNP) simulation results. The science application, operational requirements, and performance objectives for this heavy ion source will be presented.

Spherical time-encoded radiation imaging simulations

Radiation source localization is important for nuclear nonproliferation and can be obtained using time-encoded imaging systems with unsegmented detectors. A scintillation crystal can be used with a moving coded-aperture mask to vary the detected count rate produced from radiation sources in the far field. The modulation of observed counts over time can be used to reconstruct an image with the known coded-aperture mask pattern. Current time-encoded imaging systems incorporate cylindrical coded-aperture masks and have limits to their fully coded imaging field-of-view. This work focuses on expanding the field-of-view to 4π by using a novel spherical coded-aperture mask. A regular icosahedron is used to approximate a spherical mask. This icosahedron consists of 20 equilateral triangles; the faces of which are each subdivided into four equilateral triangle-shaped voxels which are then projected onto a spherical surface, creating an 80-voxel coded-aperture mask. These polygonal voxels can be made from high-Z materials for gamma-ray modulation and/or low-Z materials for neutron modulation. In this work, we present Monte Carlo N-Particle (MCNP) simulations and simple models programmed in Mathematica to explore image reconstruction capabilities of this 80-voxel coded-aperture mask.

On β-titanium alloys: tailoring gamma-ray, and neutron transmission properties for nuclear and biomedical applications

This study explores the gamma-ray and neutron transmission properties of β-Titanium alloys pivotal for their performance in nuclear and biomedical applications. Utilizing the Monte Carlo N-Particle (MCNP 6.3) simulations, we analyzed a spectrum of Ti-based alloys modified with elements like molybdenum (Mo), zirconium (Zr), niobium (Nb), and hafnium (Hf) to determine their radiation attenuation properties. Key parameters such as mass and linear attenuation coefficients, half-value layers, exposure buildup factors, and fast neutron effective removal cross-section values were computed, revealing significant enhancements in attenuation with the addition of high-Z elements. Specifically, alloys like Ti50Hf50 and (TiZr)40Cu60 exhibited superior photon and fast neutron attenuation due to their high-Z constituents. For instance, Ti50Hf50 showed a mass attenuation coefficient of 0.217 cm2/g and a half-value layer of 2.97 cm at 0.1 MeV photon energy, while (TiZr)40Cu60 demonstrated similar performance with a mass attenuation coefficient of 0.198 cm2/g and a half-value layer of 3.26 cm. These alloys also exhibited effective neutron removal cross-section values of 0.115 cm−1 and 0.130 cm−1, respectively. Alloys with lower-Z elements showed less attenuation, which may be beneficial in scenarios requiring reduced radiographic contrast in biomedical applications. The MCNP outcomes were in strong agreement with standard data, affirming the accuracy of computational methods in predicting material behavior. In conclusion, tailored alloy development is crucial for improving radiation shielding and diagnostic visualization, with broader implications for patient safety and treatment efficacy.

A Preliminary Study on Evaluation of Time-Dependent Radionuclide Removal Performance Using Artificial Intelligence for Biological Adsorbents

Background: Recently, biological adsorbents have been developed for removing radionuclides from radioactive liquid waste due to their high selectivity, eco-friendliness, and renewability. However, since they can be damaged by radiation in radioactive waste, a method for estimating the bio-adsorbent performance as a time should consider the radiation damages in terms of their renewability. This paper aims to develop a simulation method that applies a deep learning technique to rapidly and accurately estimate the adsorption performance of bio-adsorbents when inserted into liquid radioactive waste.Materials and Methods: A model that describes various interactions between a bio-adsorbent and liquid has been constructed using numerical methods to estimate the adsorption capacity of the bio-adsorbent. To generate datasets for machine learning, Monte Carlo N-Particle (MCNP) simulations were conducted while considering radioactive concentrations in the adsorbent column.Results and Discussion: Compared with the result of the conventional method, the proposed method indicates that the accuracy is in good agreement, within 0.99% and 0.06% for the R<sup>2</sup> score and mean absolute percentage error, respectively. Furthermore, the estimation speed is improved by over 30 times.Conclusion: Note that an artificial neural network can rapidly and accurately estimate the survival rate of a bio-adsorbent from radiation ionization compared with the MCNP simulation and can determine if the bio-adsorbents are reusable.

Measurement on the neutron and gamma radiation shielding performance of boron-doped titanium alloy Ti50Cu30Zr15B5 via arc melting technique

The significance of radiation shielding is on the rise due to the expanding areas exposed to radiation emissions. Consequently, there is a critical need to develop metal alloys and composites that exhibit excellent capabilities in absorbing neutron and gamma rays for effective radiation shielding. Low-density Ti-based alloys with controlled structural properties can be used for radiation protection purposes. The present research investigates boron-doped Ti-based alloy, Ti50Cu30Zr15B5, which is synthesized by arc melting technique, and its structural, mechanical properties, neutron, and gamma-ray transmission rate were investigated. Monte Carlo N-Particle simulation (MCNP6.2) code is used for calculating the Thermal (2.53 × 10−8 MeV) and fast (2 MeV) neutron transmission ratio (I/I0) dependent on the sample thickness. The Phy-x program is employed for calculating the gamma-ray LAC, MAC, HVL, TVL, and MFP values. The calculated neutron shielding performance parameters of Ti50Cu30Zr15B5 alloy were compared with materials in the literature. It was found that Ti50Cu30Zr15B5 alloy demonstrated impressive physical characteristics, suggesting that it can serve as a promising alloy for neutron and gamma-ray shielding applications.

A gamma spectroscopy-based non-destructive approach for pebble bed reactor safeguards

A gamma radiation spectroscopy study was carried out for the nuclear safeguards monitoring of fuel burnup and special nuclear material (SNM) mass in the irradiated fuel of a pebble bed reactor (PBR). The concentration of radioactive materials in irradiated PBR fuel pebbles was estimated as a function of fuel burnup for varying 235U enrichments using the Monte Carlo N-Particle (MCNP) neutronics simulation code. Gamma radiation spectra corresponding to various pebble fuel burnups were then created using MCNP modeling and simulations of a high-purity germanium (HPGe) detector. Analysis of the simulation results showed that the correlations such as (a) fuel burnup versus leftover 235U content, total Pu content, 137Cs content, 134Cs/137Cs ratio, and 154Eu/137Cs ratio; (b) 137Cs mass versus leftover 235U content, total Pu content, and fissile Pu content, and (c) 134Cs/137Cs or 154Eu/137Cs ratio versus total Pu content are sufficient to estimate pebble fuel burnup and the SNM mass.

225Ac]Ac- and [111In]In-DOTA-trastuzumab theranostic pair: cellular dosimetry and cytotoxicity in vitro and tumour and normal tissue uptake in vivo in NRG mice with HER2-positive human breast cancer xenografts

BackgroundTrastuzumab (Herceptin) has improved the outcome for patients with HER2-positive breast cancer (BC) but brain metastases (BM) remain a challenge due to poor uptake of trastuzumab into the brain. Radioimmunotherapy (RIT) with trastuzumab labeled with α-particle emitting, 225Ac may overcome this challenge by increasing the cytotoxic potency on HER2-positive BC cells. Our first aim was to synthesize and characterize [111In]In-DOTA-trastuzumab and [225Ac]Ac-DOTA-trastuzumab as a theranostic pair for imaging and RIT of HER2-positive BC, respectively. A second aim was to estimate the cellular dosimetry of [225Ac]Ac-DOTA-trastuzumab and determine its cytotoxicity in vitro on HER2-positive BC cells. A third aim was to study the tumour and normal tissue uptake of [225Ac]Ac-DOTA-trastuzumab using [111In]In-DOTA-trastuzumab as a radiotracer in vivo in NRG mice with s.c. 164/8-1B/H2N.luc+ human BC tumours that metastasize to the brain.ResultsTrastuzumab was conjugated to 12.7 ± 1.2 DOTA chelators and labeled with 111In or 225Ac. [111In]In-DOTA-trastuzumab exhibited high affinity specific binding to HER2-positive SK-BR-3 human BC cells (KD = 1.2 ± 0.3 × 10–8 mol/L). Treatment with [225Ac]Ac-DOTA-trastuzumab decreased the surviving fraction (SF) of SK-BR-3 cells dependent on the specific activity (SA) with SF < 0.001 at SA = 0.74 kBq/µg. No surviving colonies were noted at SA = 1.10 kBq/µg or 1.665 kBq/µg. Multiple DNA double-strand breaks (DSBs) were detected in SK-BR-3 cells exposed to [225Ac]Ac-DOTA-trastuzumab by γ-H2AX immunofluorescence microscopy. The time-integrated activity of [111In]In-DOTA-trastuzumab in SK-BR-3 cells was measured and used to estimate the absorbed doses from [225Ac]Ac-DOTA-trastuzumab by Monte Carlo N-Particle simulation for correlation with the SF. The dose required to decrease the SF of SK-BR-3 cells to 0.10 (D10) was 1.10 Gy. Based on the D10 reported for γ-irradiation of SK-BR-3 cells, we estimate that the relative biological effectiveness of the α-particles emitted by 225Ac is 4.4. Biodistribution studies in NRG mice with s.c. 164/8-1B/H2N.luc+ human BC tumours at 48 h post-coinjection of [111In]In-DOTA-trastuzumab and [225Ac]Ac-DOTA-trastuzumab revealed HER2-specific tumour uptake (10.6 ± 0.6% ID/g) but spleen uptake was high (28.9 ± 7.4% ID/g). Tumours were well-visualized by SPECT/CT imaging using [111In]In-DOTA-trastuzumab.ConclusionWe conclude that [225Ac]Ac-DOTA-trastuzumab exhibited potent and HER2-specific cytotoxicity on SK-BR-3 cells in vitro and HER2-specific uptake in s.c. 164/8-1B/H2N.luc+ human BC tumours in NRG mice, and these tumours were imaged by SPECT/CT with [111In]In-DOTA-trastuzumab. These results are promising for combining [111In]In-DOTA-trastuzumab and [225Ac]Ac-DOTA-trastuzumab as a theranostic pair for imaging and RIT of HER2-positive BC.

Radiation dose evaluation to organs in neonatal patients by field size during potable X-ray examination in incubators: A Monte Carlo simulation study.

Neonatal patients located in incubators are exposed to as many as 159 radiographs until discharge. To reduce the dose exposed to the patient, factors that may cause unnecessary exposure to the patient were judged. When conducting portable X-rays of neonatal patients located in an incubator, it is not easy to determine the exact field size because collimation light is exposed on the acrylic plate, an incubator canopy, and the resulting shadow is reflected on the patient's body. This study aims to measure the organ dose exposed to the patient according to the field size when a portable radiograph is given to a neonatal patient in a neonatal intensive care unit (NICU) incubator. To identify the absorbed organ dose depending on the radiation field size during portable X-ray examination of neonatal patient, a Monte Carlo N-Particle (MCNP) simulation, a SpeckCalc program, and a neonatal phantom from the ICRP 89 are applied for the calculation. According to the minimal field size (MinFS) standards of the European Commission (EC), the smaller field size is intended to measure tightly from the top of the lung apices to the bottom of the genitals; a larger field size is also calculated by adding 6 cm in width and length. Compared to the hospital C condition from the previous study, the larger and smaller field sizes are decreased by an average of 45% and 67%, respectively. Study results also show a 42% reduction in smaller field size compared to the larger field size. When taking chest and abdomen radiographic images of neonatal patients in incubators, appropriate field sizes are required to prevent inappropriate dose absorption for non-thoracic organs.

Electron, Photon, and Neutron Dose Conversion Coefficients of Lens and Non-Lens Tissues Using a Multi-Tissue Eye Model to Assess Risk of Cataracts and Retinitis.

Previous publications describe the estimation of the dose from ionizing radiation to the whole lens or parts of it but have not considered other eye tissues that are implicated in cataract development; this is especially critical for low-dose, low-ionizing-density exposures. A recent review of the biological mechanisms of radiation-induced cataracts showed that lenticular oxidative stress can be increased by inflammation and vascular damage to non-lens tissues in the eye. Also, the radiation oxygen effect indicates different radiosensitivities for the vascular retina and the severely hypoxic lens. Therefore, this study uses the Monte Carlo N-Particle simulations to quantify dose conversion coefficients for several eye tissues for incident antero-posterior exposure to electrons, photons, and neutrons (and the tertiary electron component of neutron exposure). A stylized, multi-tissue eye model was developed by modifying a model by Behrens etal. (2009) to include the retina, uvea, sclera, and lens epithelial cell populations. Electron exposures were simulated as a single eye, whereas photon and neutron exposures were simulated employing two eyes embedded in the ADAM-EVA phantom. For electrons and photons, dose conversion coefficients are highest for either anterior tissues for low-energy incident particles or posterior tissues for high-energy incident particles. Neutron dose conversion coefficients generally increase with increasing incident energy for all tissues. The ratio of the absorbed dose delivered to each tissue to the absorbed dose delivered to the whole lens demonstrated the considerable deviation of non-lens tissue doses from lens doses, depending on particle type and its energy. These simulations demonstrate that there are large variations in the dose to various ocular tissues depending on the incident radiation dose coefficients; this large variation will potentially impact cataract development.

Towards efficient Monte Carlo N-Particle simulation of a positron emission tomography (PET) via source volume definition

Monte Carlo N-Particle (MCNP) simulation has been extensively proven in nuclear medicine imaging systems, most notably in designing and optimizing new medical imaging tools. It enables more complicated geometries and the simulation of particles passing through and interacting with materials. However, a relatively long simulation time is a drawback of Monte Carlo simulation, mainly when complex geometry exists. The current study presents an alternative variance reduction technique for a modeled positron emission tomography (PET) camera by reducing the height of the source volume definition while maintaining the geometry of the simulated model. The National Electrical Manufacturers Association (NEMA) of the International Electrotechnical Commission (IEC) PET's phantom was used with a 1 cm diameter and 7 cm height of line source placed in the middle. The first geometry was fully filled the line source with 0.50 mCi radioactivity. In contrast, the second geometry decreased the source definition to 2.4 cm in height, covering 1 cm above and below the sub-block detector level. The source volume definition approach led to a 71% reduction in the total photons to be simulated. Results showed that the proposed variance reduction strategy could produce spatial resolution as precise as fully filled geometry and sped up the simulation time by approximately 65%. Hence, this strategy can be utilized for further PET optimizing simulation studies.

Calculation of Spatial Response of a Collimated Segmented HPGe Detector for Gamma Emission Tomography by MCNP Simulations

We have proposed a planar electronically segmented high-purity germanium (HPGe) detector concept in combination with a multislit collimator for gamma emission tomography. In this work, the spatial resolution achievable using the collimated segmented HPGe detector was evaluated, prior to the manufacture and operation of the detector. The spatial response of a collimated segmented HPGe detector concept was evaluated using simulations performed with Monte Carlo N-Particle (MCNP) transport code MCNP6. The full detector and multislit collimator system were modeled, and for the quantification of the spatial response, the modulation transfer function (MTF) was chosen as a performance metric. The MTF curve was obtained through the calculation of the line spread function (LSF) by analyzing the simulated projection data. In addition, tomographic reconstructions of the simulated simplified test objects were made to demonstrate the performance of the segmented HPGe detector in planned application. For 662-keV photons, the spatial resolution obtained was approximately the same as the collimator slit width for both the 100- and 150-mm-long collimators. The corresponding spatial resolution at 1596-keV photon energy was almost twice the slit width for the 100-mm collimator, due to the partial penetration of the high-energy gamma rays through the collimator bulk. For a 150-mm-long collimator, an improved resolution was obtained.

Evaluation of radiation dose to organs of neonatal patients during portable X-ray examination in incubators: A Monte Carlo simulation study.

Infants admitted to neonate intensive care units (NICUs) are placed in incubators to maintain body temperature and condition, which undergo normal radiographs and are exposed to radiation. Furthermore, different incubator structures in different hospitals exhibit varying object to image receptor distance (OID), source to image receptor distance (SID), presence of canopy, which results in variations in X-ray radiation conditions and doses absorbed by the neonatal patients. To measure organ dose exposed to neonatal patient in different incubator settings. A portable X-ray was performed on a neonatal patient placed in an incubator to identify disease progress, the injection path of the drug, and various factors. To minimize direct contact between neonatal patients and image receptor, radiologic technologists place the image receptor on a tray underneath the incubator and place the portable X-ray tube on top of the acrylic canopy of the incubators. SID and OID settings and value of organ dose exposed to the patient varied based on the incubator structure, and the organ absorbed dose was determined using Monte Carlo N-Particle (MCNP) simulation, PC-based Monte Carlo program (PCXMC) 2.0 simulation, and neonate phantoms. Evaluations of organ dose of neonatal patients in three hospitals with different incubator settings reveal that the average organ dose differs by 36% depending on change in OID and SID settings and reduces by 10% with an acrylic canopy. Therefore, owing to the presence of an acrylic canopy on the top of the incubator and the longer SID with the corresponding shorter OID, a lower dose was absorbed by organs of neonatal patient. Our results provide proof that proper incubator standard decreases organ dose to neonatal patient during continuously diagnostic X-ray procedure.

The Effect of Mouse Size on Dose from an X-Rad320 Irradiator.

Irradiation protocols for murine experiments often use standardized dose rate estimates for calculating dose delivered, regardless of physical variations between mouse subjects. This work sought to determine the significance of mouse size on absorbed dose. Five mouse-like phantoms of various sizes based on the mouse whole-body (MOBY) model were 3D printed. The phantoms were placed in an X-Rad320 biological irradiator and a standard irradiation protocol was used to deliver dose. Dose was measured using thermoluminescent dosimeter (TLD) microcubes inside each phantom, and the relative readings were used to calculate output factors (OFs), normalized to the phantom of median volume. Additionally, the OF for each mouse was simulated in Monte Carlo N-Particle (MCNP) code. For both the TLD measurements and MCNP simulations, the OF for each mouse was determined by both experiments and calculations to be unity within the relative standard uncertainties (k = 1). This work supports comparing results across various studies using the X-Rad320 irradiator without need for corrections based on mouse size.

Experimentally validated x-ray image simulations of 50 μm x-ray PIV tracer particles

We evaluate Beer–Lambert (BL) ray-tracing and Monte Carlo N-Particle (MCNP) photon tracking simulations for prediction and comparison of x-ray imaging system performance. These simulation tools can aid the methodical design of laboratory-scale x-ray particle image velocimetry (XPIV) experiments and tracer particles by predicting image quality. Particle image signal-to-noise ratio (SNR) is used as the metric of system performance. Simulated and experiment data of hollow, silver-coated, glass sphere tracer particles (AGSF-33) are compared. As predicted by the simulations, the AGSF-33 particles are visible with a SNR greater than unity in 100 ms exposure time images, demonstrating their potential as x-ray PIV or particle tracking velocimetry (XPTV) tracers. The BL approach predicts the image contrast, is computationally inexpensive, and enables the exploration of a vast parameter space for system design. MCNP simulations, on the other hand, predict experiment images slightly more accurately, but are more than an order of magnitude more computationally expensive than BL simulations. For most practical XPIV system design applications, the higher computational expense of MCNP is likely not justified by the modest accuracy improvement compared to BL.

Pedikül Vidası İçin Kütle Zayıflama Katsayılarının Teorik ve Monte Carlo Simülasyon Teknikleri ile Hesaplanması

Spine fixation is required in cases such as congenital spinal curvatures, vertebral fractures, sagittal collapse over time, painful kyphosis, and bone load due to tumors. Although there are many methods in the literature, the most commonly used spine fixation method is the fixation with pedicle screws. In these cases, it is known that pedicle screws are used frequently in the body. In this study, how the radiological exposure of the pedicle screws in the vertebral column that dose was evaluated by simulation methods. First, the elemental analysis of the pedicle screw was analyzed via Scanning Electron Microscopy (SEM) equipped with the Energy Dispersive X-ray Spectroscopy (EDS). Then, the elemental compositions of the pedicle screw obtained were used for simulation codes. subsequently, the half-value thickness and the attenuation coefficient calculations were conducted for the pedicle screw and vertebral column. Both XCOM software and MCNP (Monte Carlo N-Particle) simulation code were used to obtain photon interaction parameters within the energy range of 60-250 keV.

Development of Scan Survey Adjust Method for Radioactively Contaminated Sediment in Underwater Conditions

Kim, J.-H.; Cho, G.-S.; Lee, J.-J.; Hong, S.-B.; Lee, E.-J.; Lee, B.-C., and Seo, B.-K., 2021. Development of scan survey adjust method for radioactively contaminated sediment in underwater conditions. In: Lee, J.L.; Suh, K.-S.; Lee, B.; Shin, S., and Lee, J. (eds.), Crisis and Integrated Management for Coastal and Marine Safety. Journal of Coastal Research, Special Issue No. 114, pp. 325–329. Coconut Creek (Florida), ISSN 0749-0208. The nuclear power plants located near the coast and river to supply the cooling water. Therefore, unexpected radioactive materials leakage can contaminate the sediment on the coast, river, and lake. The conventional hotspot assessment methods are sampling and scan survey. The sampling method has limitations due to the sample representation, while the scan survey method is an easy way to assess hotspot distribution. However, the scan survey method can overestimate or underestimate the hotspot size depending on the radioactivity concentration, because this method determines the hotspot is exist when the measured count rate or dose rate exceeds the criteria level. In this study, the scan survey adjust method of sediment scan survey has developed. To this, the distribution of spatial effective dose rate was assessed by MCNP(Monte Carlo N-Particle) simulation. Also, the overestimation factor, which is the ratio of the actual hotspot size and estimated hotspot size was calculated. For the scan survey adjust method devised using the overestimation factor, maximum dose rate, and dose rate increase factor. Using this method, the scan survey experiment results were adjusted. As a result, 137Cs survey result that was underestimated were adjusted 11.51∼73.05% overestimate. 60Co survey result that was 2∼3 times overestimated was adjusted 2.02∼45.73% overestimate. Therefore, the scan survey adjusts method can prevent hotspot underestimation and reduced the relative error. This method can improve the accuracy of the hotspot scan survey in sediment.

Preliminary Studies of Perovskite-Loaded Plastic Scintillator Prototypes for Radioactive Strontium Detection

Functional plastic scintillators have attracted much attention for their usefulness in on-site monitoring and detection in environments. In this study, we elucidated a highly reliable and functional plastic scintillator for detection of radioactive strontium, which means a potent perovskite-loaded polymeric scintillation material based on epoxy and 2,5-diphenyloxazole (PPO). Moreover, Monte Carlo N-Particle (MCNP) simulation was performed to optimize the thickness of a plastic scintillator for efficient strontium detection. A thickness of 2 mm was found to be the optimum thickness for strontium beta-ray detection. A newly developed plastic scintillator with 430 nm emission from perovskite loading could trigger scintillation enhancement employing potential indication of perovskite energy transfer into a photomultiplier (PMT) detector. Furthermore, the response to beta-ray emitter of 90Sr was compared to commercial scintillator of BC-400 by exhibiting detection efficiency in the energy spectrum with a fabricated perovskite-loaded plastic scintillator. We believe that this suggested functional plastic scintillator could be employed as a radiation detector for strontium detection in a wide range of applications including decommissioning sites in nuclear facilities, nuclear security and monitoring, nonproliferation, and safeguards.

MCNP modelling of radiation effects of the Dragonfly mission's RTG on Titan

The Dragonfly mission to Titan is proposed to use a radioisotope thermoelectric generator to supply electrical and thermal power to the spacecraft during flight and once it arrives at its destination. As no spacecraft continuously powered by an RTG has been flown down through Titan's atmosphere before, the effects of such a generator's radiation field on the surrounding environment are unknown. MCNP (Monte Carlo N-Particle) simulations were run on a simplified model of the generator in its protective hull section in a coarse-grained Titan environment with both neutron and photon transport modelled, the former to determine the degree to which the surface was likely to become activated, and the latter to determine the energy likely to be deposited in a surface covering of tholin-analogue hydrocarbons, and consequently the potential for damage to chemical bonds found therein. Tests on the model also provided data which allowed for a determination of the degree to which the insulation surrounding the generator spacecraft might be activated by stray neutrons, how much heating might be expected in the hull section surrounding the generator and how much attenuation the beam from the spacecraft's pulsed neutron generator is likely to experience due to Titan's thick nitrogen atmosphere.