Investigation of Rhodnius prolixus head anatomy using microtomography.

In the last years, microtomography has proved to be a powerful technique on insects' studies, allowing a detailed view of the structures' internal with a high resolution. One of the most important advantages about the use of microtomography in these studies is the fact that the dissection is not necessary, which decreases considerably the number of samples used on the insects' research. Some insects are used constantly in studies about morphology, metamorphosis, and reproduction, because they work as a model for others, and Rhodnius prolixus is one of the most studied in this group. This insect is also one of the main insect vectors of Chagas disease that kills around 12,000 people every year in Latin America. Some studies using laboratory microtomography conventional scanners combining with the correct staining methods have proved that it could be a powerful tool in biological research, allowing the visualisation of low-density tissues. The main goal of the present work was to use staining protocols to study Rhodnius prolixus with laboratory microtomography conventional scanners. The experiments were carried out at the imaging lab in the Theoretical Biology Department, University of Vienna, using an Xradia MicroXCT and at the University of Oslo, using a Skyscan 2211.

Fast radiological safety evaluation applied to maintenance in cargo and container inspection facilities.

The application of nuclear technologies in a cargo and container inspection facility can increase the risk of accidents. Estimating the radiation dose in the controlled area generates critical information for elaborating routines aimed at establishing more effective safety procedures. For radiological protection purposes, mapping ambient dose equivalent H*(10) levels is crucial. The radiation source used was a fixed linear accelerator of 4.5 MeV. Five RadEye PRD-ER (Thermo Fisher Scientific) personal radiation monitors and five Geiger-Müller MRAD 111 (Ultra Radac) personal radiation monitors were used for the radiation measurements. The highest ambient equivalent dose rate and dose per scan were found with the Geiger-Müller monitors at values of 5.76E-01 mSv/h and 1.12E-03 mSv, respectively. The results showed that for public individuals, the number of scans at the point of highest dose rate value cannot exceed 893-unit operations. Additionally, the risks involved in the abnormal situation (increased H*(10)) were estimated by using a model to predict the development of solid cancer as a result of occupational radiological exposure. This procedure highlights the risks involved, hence providing initial support to the decision process.

Radiometric analysis of Brazilian commercial ornamental stones.

The term "commercial ornamental stones" comprises different natural stones with different mineralogical components and the distribution and amount of natural radionuclides depended on the mineralogy. Few studies on natural radioactivity in ornamental stones carried out in Rio de Janeiro (Brazil) and the present work were carried out in order to analyze the distribution of activity concentration in these materials. The activity concentration of 238U, 235U, 226Ra, 232Th and 40K in Ornamental Stones measured using γ-ray spectrometry to estimate the radiation hazard as well as establishing a database for radioactivity levels. Samples were collected in stores in Rio de Janeiro, carefully transported, weighed, packed with PVC film and left to rest for at least 45 days to reach secular equilibrium. Samples analyzed by gamma spectrometry using a hyper pure germanium detector (HPGE). Detection efficiency for each sample was obtained using LabSOCS software. The activity concentrations for samples were within the ranges (0.83 ± 0.15 to 39.96 ± 1.19, 8.04 ± 2.54 to 111.36 ± 13.33, 0.43 ± 0.03 to 264.58 ± 0.98, BDL to 8.17± 0.65 and 1.30 ± 0.57 to 1567.01 ± 65.08) Bq⋅kg-1 for 238U, 226Ra, 232Th, 235U and 40K, respectively. The world average for building materials is 50 Bq⋅kg-1 for 238U, 50 Bq.kg-1 for 232Th and 500 Bq⋅kg-1 for 40K. In the present work, the average for granite was 19.43 ± 0.39 Bq⋅kg-1 for 238U, 56.70 ± 1.13 Bq⋅kg-1 for 232Th and 1113.28 ± 20.27 Bq⋅kg-1 for 40K, thus the values for 232Th and 238U are above the world average. Analyzing for the granite that showed the highest concentrations of activity the average values of Transport Rate was 3.12 ± 0.04 Bq⋅m-2⋅s-1, Emanation Rate was 0.16 ± 0.00 Bq⋅kg-1⋅s-1, Exhalation Flow was 530.26 ± 6.05 Bq⋅m-3, Effective Equivalent Dose Due to Radon Exposure was 0.78 ± 0.01 WLM⋅year-1 and Annual Effective Dose Due to Radon and Thoron was 7.02 ± 0.10 mSv⋅year-1. Annual Effective Dose Due to Radon and Thoron it is above the annual limit for public individuals which is 1 mSv⋅year-1 and below the occupationally exposed individual, which is 20 mSv⋅year-1 in an average of 5 years, it does not exceed the maximum level recommended by ICRP-60.

Excess lifetime cancer risk due to ingestion of fish from a dam reservoir associated with a nuclear facility in Brazil.

Fish from a Funil dam reservoir associated with a Nuclear Fuel Factory were sampled aiming to assess the radiological risk due to ingestion. Funil dam reservoir is a strategic site, once it receives effluent from the industrial complex that performs isotopic enrichment of uranium and conversion of UF6. The mean activity concentrations obtained for 40K, 226Ra, 228Ra and 228Th were 57.81, 0.41, 0.92 and 0.49 Bq·kg-1, respectively. Lifetime cancer risk was estimated in ∼10-5 and no action needs to be taken.

Application of radiological assessment as water quality criterion for effluent release in a Brazilian uranium mine.

Uranium mining causes several radiological impacts on the surrounding environment, notably in the water bodies, mainly due to the release of long half-life radionuclides from the 238U and 232Th series. The Ore Treatment Unit, an old uranium mine undergoing decommissioning, has three points of liquid effluent release (#014, #025, and #076). For current study, 78 samples of water were collected at #014, 33 samples at #025, and 63 samples at #076. The radionuclides were analyzed by gross alpha count, gross beta count, and by arsenazo spectrophotometry. Analyses were carried out using the radiological water quality criterion established by World Health Organization and other organizations, together with the Brazilian legislation, to assess if the released effluents may be used unrestrictedly by the individuals of the public. At #014, the mean values of activity concentration (AC), in Bq·L-1, were as follows: Unat = 0.107, 226Ra = 0.035, 210Pb = 0.031, 232Th = 0.007, and 228Ra = 0.049. At #025 the mean values of AC, in Bq·L-1, were as follows: Unat = 0.086, 226Ra = 0.015, 210Pb = 0.028, 232Th = 0.006, and 228Ra = 0.032. Finally, at point #076, the mean AC values, in Bq·L-1, were as follows: Unat = 3.624, 226Ra = 0.074, 210Pb = 0.054, 232Th = 0.013, and 228Ra = 0.069. The current study showed that natural radionuclides were not in secular equilibrium. Despite uranium presented its values outside the limits of guidance levels, it can be state that the unrestricted use of effluents released in the three water bodies is authorized from the radiological point of view. In terms of dose rate, the releases at three points were within the radiological limits of potability. On the other hand, in an additional analysis, #76 presented chemical toxicity above the authorized value, pointing the need of restricted use of water from the point of view of chemical toxicity.

Relationships between natural radionuclides activity concentration in soils of the State of Alagoas, Brazil and WRB/FAO soil classification, types of parental rocks and rainfall variability.

Climate, soil type and geology that can influence the activity concentrations of natural radionuclides in soils. For this, 41 samples were collected in the first 25 cm of soil from the state of Alagoas, Northeast Brazil. This sample collection represents the mains soils classifications, parental rocks, and rainfall variability in this locality. The samples of soils were processed and analyzed by means of gamma-ray spectrometry for the 40K, 226Ra and 228Ra. The non-parametric Mann-Whitney U test was applied to perform the statistical inference on the differences in activity concentration for each soil class studied. There was a significant difference in the 40K activity distributions between the combination for most soil types. The results for 226Ra show no significant difference. However, Planosols differ significantly from Yellow Acrisols and Regosols for 228Ra. The strong relationship of the 40K concentration with the rock classes was shown. On the other hand, an association of the activity concentration of 226Ra and 228Ra with the rock type was not evidenced. There was a negative correlation between the 40K activity concentration and the rainfall index. 226/228Ra concentrations showed no correlation with the rainfall index.

S-values for radium-223 and absorbed doses estimates for 223RACL2 using three computational phantoms.

Radium-223 dichloride (223RaCl2), approved by FDA (Food and Drug Administration) in 2013 and in Brazil by ANVISA (Agência Nacional de Vigilância Sanitária) in 2016, offers a new therapeutic option for bone metastases from castration-resistant prostate cancer (CRPC). The advantages of radionuclide therapy for bone metastases include the simultaneous treatment of multiple lesions at the same time. The activity prescription is based on the patient's body weight, disregarding the absorbed dose limit of 2 Gy in the organ at risk: bone marrow. This study focuses on Internal Dosimetry for 223RaCl2 therapy aiming to apply biokinetic models described in the literature to estimate absorbed doses in the organs of interests, especially for the bone marrow. For this purpose, the present paper compares and validates the GATE Monte Carlo simulation with the Radioactive Decay Module (RDM) and calculates a set of S-values for Radium-223 radionuclide using male and female XCAT computational models. Moreover, a comparison of S-values for Radium-223 for three male computational models with different anatomies is also evaluated, Male (standard), Pat1 (lower body weight) and Pat2 (highest body weight). A comprehensive set of S-values was calculated for the Male model, 30 source-regions and 47 target-regions, and for Female model, 30 source-regions and 42 target-regions for Radium-223 and its decay scheme: Radon-219, Polonium-215, Lead-211, Bismuth- 211, Polonium-211 and Thallium-207. The new set of S-values will facilitate absorbed dose calculations for Radium-223 therapy. In addition, Absorbed Dose Evaluation for 223RaCl2 therapy was estimated for three different biodistributions described in the literature within three male computational models. For all biodistributions, the Pat2 phantom has a greatest absorbed dose within the red marrow, when compared with Male and Pat1.

Assessment of released natural radionuclides by waste rock pile and mining pit associated with a uranium mine at Caldas, Minas Gerais, Brazil.

The Ore Treatment Unit was a uranium mining company that is currently being decommissioned. The local rainfall index makes it necessary to release effluents into the environment. After releasing, the wastewater is available for unrestricted use. Current study aims to use national and international recommendations to assess the radiological potability of released effluents at one of the three points of company's interface with the environment. Twenty-four samples of water were collected and activity concentrations (AC) were obtained by gross alpha count, gross beta count, and for arsenazo spectrophotometry. Statistical analysis techniques were applied to the data with the purpose of understanding the results for the soluble, particulate, and total fractions. The mean AC for effluents were 3.580, 0.082, 0.103, 0.063, and 0.090 Bq L-1 for Unat, 226Ra, 210Pb, 232Th, and 228Ra, respectively, for the total fraction. The analysis of variance pointed to Unat as a critical radionuclide, since it presented more than 90% of the total AC released into the environment. Pearson's R2 pointed to soluble fraction as a major contributor to the total AC released. The guidance level proposed by WHO was used to assess the radiological potability of the effluents. The results obtained indicated the need for trigger other analyses. Committed effective dose was estimated due to the unrestricted use of effluents and the value obtained, 0.23 mSv year-1, was below the maximum allowed limit. Finally, the radiotoxicity of the released effluent was evaluated and the value obtained was ~ 50% of the maximum allowed limit. In conclusion, the present study showed that the level of radioactivity released into the environment by the Ore Treatment Unit does not present a radiological risk to the surrounding population.

Absorbed dose rate for marine biota due to the oil spilled using ICRP reference animal and Monte Carlo simulation.

The current study aimed to obtain dose conversion coefficients for marine animals due to an oil spill accident using two variables: crude oil activity concentration and organism depth. Thorium series presented a dose contribution twice that uranium series for similar conditions. Bi-214 and Tl-208 stood out for delivering a higher dose rate for uranium and thorium series, respectively. Results obtained can be used to assess the maximum exposure time for emergency oil control, removal, and mitigation in an oil spill accident.

Impact of the affected population size assessment on the decision-making after a nuclear event.

In this study an improvised nuclear device (IND) is simulated using a software called HotSpot. The explosion took place in a theoretical central business district (CBD), for the major issue addressed in this paper is the comparison of two methods used for estimating the size of the potentially affected population. The first method estimates the size by multiplying the local average demographic density by the area of the zone of interest. The second method uses the population density gradient model developed by Colin Clark in 1951. The comparison of the two methods enables authorities to better estimate the allocation of resources. The conservative approach allocates the maximum resources possible. However, the Clark model enables a more realistic approach which allocates minimum resources to the emergency response. This study shows how accurate information can be quintessential for authorities to maximize the efficiency of their decisions.

Committed effective dose and lifetime cancer risk due to ingestion of natural radionuclides in grains grown in an area of high background radiation.

The current study aimed at estimating committed effective dose and cancer risk due to the intake of K-40, Ra-226, Ra-228 and Th-228 present in grains grown in an HBRA. The highest activity concentrations found were (606.2 ± 25.13), (8.07 ± 6.37), (10.01 ± 1.45), (43.97 ± 5.54) Bq.kg-1 for K-40, Ra-226, Ra-228 and Th-228, respectively. The committed effective dose estimated was 0.5 mSv.y-1, and the estimated cancer risk suggested that uninterrupted and unrestricted consumption of beans grown in this HBRA is not desirable.

Lifetime cancer risk increase due to consumption of some foods from a High Background Radiation Area.

Activity concentration (AC) in foods produced and commonly consumed in a High Background Radiation Area (HBRA) was analyzed. The AC were obtained by spectrophotometry and by the radiochemical separation method. The AC were up to 104 times higher than the AC for both UNSCEAR suggested values and non-HBRA. It was noted that the lifetime cancer risk was increased in 4 decimal places, taking the risk from the "statistically negligible range" (<10-6) to "middle range" (between 10-4 and 10-6).

MCNPX computational modeling applied to the potential dose rates calculation of cargo scanning.

This study focusses on the risk of potential exposure to radiation for personnel driving a truck as well as illegal individuals being transported in cargo containers. Inspection facilities usually use a high energy linear accelerator (linac) in order to inspect the cargo. Since this type of equipment has associated health risks due to potential unwanted exposure, the occupational and public dose limits should be calculated in order to develop safer work conditions. This work used a computation model running the code MCNPX to simulate a typical cargo inspection facility which used a linac operating at 4.5 MeV. Two scenarios were considered: (1) exposure of the driver to the primary beam due to a potential failure of the safety sensors; and (2) dose received by an illegal individual being transported inside the cargo container. The results show a dose of 0.8514 mSv per scan for the driver exposed to the primary X-ray beam, and 0.1997 mSv per scan for an individual transported in the cargo box. In conclusion, both the individual and the driver received a dose below the acceptable limit considered safe for an individual (1 mSv/year). However, that was the value of one scan; in a case in which multiple scans would be performed, the dose limit can be quickly exceeded. In that case, the limit would be exceeded by the driver faster than by the individual in the cargo.

Comparison between codes MCNPX and Gate/Geant4 in volume fraction studies.

Knowing the volume fraction in a multiphase flow is of fundamental importance in predicting the performance of many systems and processes, it has been possible to model an experimental apparatus for volume fraction studies using Monte Carlo codes. Artificial neural networks have been applied for the recognition of the pulse height distributions in order to obtain the prediction of the volume fractions of the flow. In this sense, some researchers are unsure of which Monte Carlo code to use for volume fractions studies in two-phase flows. This work aims to model a biphasic flow (water and air) experiment in a stratified regime in two Monte Carlo-based codes (MCNP-X and Gate/Geant4), and to verify which one has the greatest benefits for researchers, focusing on volume fractions studies.

Monitoring system of oil by-products interface in pipelines using the gamma radiation attenuation.

This paper presents a methodology to precise identify the interface region, which is formed in the transport of petroleum by-products in polyducts, using gamma densitometry. The simulated geometry is compose for a collimated 137Cs source and a NaI(Tl) detector to measure the transmitted beam. The modeling was validated experimentally on stratified flow regime using water and oil. The different volume fractions were calculated using the MCNPX code in order to determine the region interface with an accuracy of 1%.

Flow regime and volume fraction identification using nuclear techniques, artificial neural networks and computational fluid dynamics.

Knowledge of the flow regime and the volume fraction in multiphase flow is of fundamental importance in predicting the performance of many systems and processes. This study is based on gamma-ray pulse height distribution pattern recognition by means of an artificial neural network. The detection system uses appropriate one narrow beam geometry, comprising a gamma-ray source and a NaI(Tl) detector. The models for annular and stratified flow regimes were developed using MCNPX code, in order to obtain adequate data sets for training and testing of the artificial neural network. Several experiments were carried out in the stratified flow regime to validate the simulated results. Finally, Ansys-CFX was used as computational fluid dynamics software to simulate two different volume fractions, which were modeled and transformed in voxels and transferred to MCNPX code. The use of computational fluid dynamics is of great importance, because it brings the studies closer to the reality. All flow regimes were correctly recognized and the volume fractions were appropriately predicted with relative errors less than 1.1%.

Urban critical infrastructure disruption after a radiological dispersive device event.

This study aims to evaluate the impacts of the activation of a hypothetical radiological dispersal device (RDD) on the urban critical infrastructure (health facilities and public transport). A densely populated urban region was chosen as a scenery. Additionally, the influence of local environmental factors in the post-detonation process was verified. The source term was Cs-137 due to its mobility in the environment and relative ease of access. The approach used for the evaluation of the consequences was a computer simulation by Gaussian modeling. The HotSpot Health Physics Codes software was applied in conjunction with the RESRAD-RDD software. The results suggest that there is a strong influence of the local atmospheric stability classes (Pasquill-Gifford classes) on both the total equivalent effective dose (TEDE) and soil contamination. Consequently, the impacts on critical urban infrastructure follow the same trend. The method used for comparing the simulated and reference limits was the proportional ratio. All calculated values for radioactive contamination were divided by the reference value adopted by the RESRAD-RDD model for urban critical infrastructure. The results indicate that the information compiled is useful to support the decision-making process, although it is not sufficient to provide care and support for longer periods than those considered in the initial response phase.

Modeling of a neutron irradiator using Monte Carlo.

The Nuclear Engineering Department of the Military Institute of Engineering (SE/7-IME) is designing and simulation a neutron irradiator with 1 Ci of 241Am-9Be source. The objective of this irradiator is to generate a flux of neutrons to be used in research and teaching maintaining, for purposes of radiological protection, the rate of ambient dose equivalent, H*(10), below 10 µSv/h at 30 cm from the surface. This paper presents the proposed irradiator, values of H*(10) at different distances from the irradiator and the neutron flux in different points of the beam irradiation, all calculated using the MCNPX code.

Shielding implications on secondary radiation doses in prostate cancer treatment.

Medical linear accelerators (linacs) require a physical structure designed to provide adequate structural support which ensures the safety of patients, operators and the general public. During a radiotherapy session, healthy tissues are exposed to radiation, even with these safety guarantees. This unwanted exposure may increase the likelihood of developing secondary cancer. This work uses the MCNP-5 code to computationally simulate a conformational 3D radiotherapy protocol for prostate cancer. Also, it investigates the potential effects of radiotherapy room shielding composition on equivalent and effective doses in the patient's body. A computational model of an actual room was developed considering a Varian Trilogy linac operating at 10 MeV. This model enabled dose calculations for an anthropomorphic phantom called REX to be performed. This phantom has sufficient details of all relevant organs and tissues needed to estimate the effective dose of the patient. The treatment protocol modeled in this study came from the database of patients treated by the Brazilian National Cancer Institute (Inca). For this protocol, the total dose to be applied to the patient is equally distributed over the four gantry inclination angles (0°, 90°, 180° and 270°). The simulated results suggested that the equivalent dose on different organs and tissues has been increased by concrete shielding. Regarding the effective dose due to the presence of additional shielding (steel or lead), the simulation suggests that such variations can be considered small. Overall the results allowed quantifying the specific contribution of concrete, lead, and steel as part of shielding on the equivalent and effective doses in the patient.

Brazilian regulatory authority contribution to the shielding dimensioning model of radiotherapy rooms proposed by the NCRP 151.

The National Council on Radiation Protection and Measurements (NCRP) Report No. 151 is an essential document for bunker design commonly applied for radiotherapy treatment rooms. This document is used as a reference by several countries, including Brazil. The objective of this study is to evaluate the shielding dimensioning methodology recommended by NCRP 151, and compare it with the one adopted by the Brazilian regulatory authority. Radiotherapy rooms and respective doors were designed to use linear accelerators operating at 6, 10, 15, and 18 MeV under two different ways: (a) applying exclusively the methodology recommended by the NCRP 151, and (b) taking into consideration the complementary recommendations from the Brazilian authorities. The results suggest that designers in Brazil can count on at least 4 and 11% safety margin for dimensioning primary barriers in controlled and free areas respectively. Also 8% for secondary barriers in controlled areas, 9.7% for secondary barriers adjacent to the primary belt of free areas, and 6.6% for the lead of the doors.