Radiological characterization of the tailings of an abandoned copper mine using a neural network and geostatistical analysis through the Co-Kriging method.

The radiological characterization of soil contaminated with natural radionuclides enables the classification of the area under investigation, the optimization of laboratory measurements, and informed decision-making on potential site remediation. Neural networks (NN) are emerging as a new candidate for performing these tasks as an alternative to conventional geostatistical tools such as Co-Kriging. This study demonstrates the implementation of a NN for estimating radiological values such as ambient dose equivalent (H*(10)), surface activity and activity concentrations of natural radionuclides present in a waste dump of a Cu mine with a high level of natural radionuclides. The results obtained using a NN were compared with those estimated by Co-Kriging. Both models reproduced field measurements equivalently as a function of spatial coordinates. Similarly, the deviations from the reference concentration values obtained in the output layer of the NN were smaller than the deviations obtained from the multiple regression analysis (MRA), as indicated by the results of the root mean square error. Finally, the method validation showed that the estimation of radiological parameters based on their spatial coordinates faithfully reproduced the affected area. The estimation of the activity concentrations was less accurate for both the NN and MRA; however, both methods gave statistically comparable results for activity concentrations obtained by gamma spectrometry (Student's t-test and Fisher's F-test).

Pixel-Grouping G(E) Functions for Estimating Dose Rates from Unknown Source Distributions with a Position-Sensitive Detector.

Estimating accurate radiation doses when a radioactive source's location is unknown can protect workers from radiation exposure. Unfortunately, depending on a detector's shape and directional response variations, conventional G(E) function can be prone to inaccurate dose estimations. Therefore, this study estimated accurate radiation doses regardless of source distributions, using the multiple G(E) function groups (i.e., pixel-grouping G(E) functions) within a position-sensitive detector (PSD), which records the response position and energy inside the detector. Investigations revealed that, compared with the conventional G(E) function when source distributions are unknown, this study's proposed pixel-grouping G(E) functions improved dose estimation accuracy by more than 1.5 times. Furthermore, although the conventional G(E) function produced substantially larger errors in certain directions or energy ranges, the proposed pixel-grouping G(E) functions estimate doses with more uniform errors at all directions and energies. Therefore, the proposed method estimates the dose with high accuracy and provides reliable results regardless of the location and energy of the source.

Validation study of ambient dose equivalent conversion coefficients for radiocaesium distributed in the ground: lessons from the Fukushima Daiichi Nuclear Power Station accident.

Ambient dose equivalent conversion coefficients (ADCRCs) for converting a radiocaesium inventory to ambient dose equivalent rates (air dose rates) depend on the vertical distribution of radiocaesium in soil. To access the validity of ADCRCs, the air dose rate at 1 m above ground and the vertical distribution of radiocaesium in the soil around the Fukushima Daiichi Nuclear Power Station (FDNPS) present between 2011 and 2019 were measured in the current study. ADCRCs were calculated using air dose rates and three different parameters representing the vertical distribution of radiocaesium in soil: (1) relaxation mass depth (ß), (2) effective relaxation mass depth (ßeff) and (3) relaxation mass depth recommended by the International Commission on Radiation Units and Measurements before the FDNPS accident (ßICRU). When ADCRCs based on ß and ßeff were compared to those based on ß and ßICRU, a positive correlation was found. To confirm the applicability of the ADCRCs based on the three types of ß values, radiocaesium inventories were estimated using the air dose rates and ADCRCs, and the obtained results were compared to the radiocaesium inventory calculated using soil sample measurements. Good agreement was observed between the radiocaesium inventories estimated using the ADCRCs based on ß and ßeff and measured by investigating soil samples. By contrast, the radiocaesium inventory estimated using the ADCRCs based on ßICRU was overestimated compared with that measured by investigating soil samples. These findings support the applicability of ADCRCs based on ß and ßeff in the Fukushima region. Furthermore, the ßICRU result suggests that differences in soil characteristics between Japan and other countries should be considered for evaluating ADCRCs.

Measurement of ambient gamma dose rate in Metro Manila, Philippines, using a portable NaI(TI) scintillation survey meter.

Radiological data such as ambient dose equivalent rate obtained from radiation monitoring in Metro Manila are useful for the detection of any anomalous increase of radiation dose rate levels due to nuclear or radiological emergencies. In this study, ambient dose equivalent rates were measured in different locations in Metro Manila using a portable NaI(Tl) scintillation survey meter to determine the background radiation levels within the capital. Ambient dose equivalent rates measured range from 32.7 ± 2.2 to 59.3 ± 8.7 nanosieverts per hour (nSv/h) with computed mean and median values of 41.7 and 39.6 nSv/h, respectively. These values were lower than the Philippines' average dose rate which is 52 nanograys per hour (nGy/h). No significant trend was also observed in the monthly variation of ambient dose equivalent rate for most locations, with the dose rates being relatively consistent throughout a year. No significant trend was further observed in the monthly variation of ambient dose equivalent rate for the whole Metro Manila. Data obtained in this study were used to develop a dose rate distribution map of Metro Manila which could be used as a baseline reference of emergency responders for environmental radioactivity monitoring during nuclear or radiological emergencies that may affect Metro Manila.

Variation of indoor radon concentration and ambient dose equivalent rate in different outdoor and indoor environments.

Subject of this study is an investigation of the variations of indoor radon concentration and ambient dose equivalent rate in outdoor and indoor environments of 40 dwellings, 31 elementary schools and five kindergartens. The buildings are located in three municipalities of two, geologically different, areas of the Republic of Macedonia. Indoor radon concentrations were measured by nuclear track detectors, deployed in the most occupied room of the building, between June 2013 and May 2014. During the deploying campaign, indoor and outdoor ambient dose equivalent rates were measured simultaneously at the same location. It appeared that the measured values varied from 22 to 990 Bq/m(3) for indoor radon concentrations, from 50 to 195 nSv/h for outdoor ambient dose equivalent rates, and from 38 to 184 nSv/h for indoor ambient dose equivalent rates. The geometric mean value of indoor to outdoor ambient dose equivalent rates was found to be 0.88, i.e. the outdoor ambient dose equivalent rates were on average higher than the indoor ambient dose equivalent rates. All measured can reasonably well be described by log-normal distributions. A detailed statistical analysis of factors which influence the measured quantities is reported.

Measurement of energy and directional distribution of neutron ambient dose equivalent for the 7Li(p,n)7Be reaction.

Double differential neutron fluence distributions were measured in the 7Li(p,n)7Be reaction for proton beam energies 7, 9 and 12 MeV. Seven liquid scintillator based detectors were employed to measure neutron fluence distributions using the Time of Flight technique. Neutron ambient dose equivalents were determined from the measured fluence distribution using ICRP (International Commission on Radiological Protection) recommended fluence to dose equivalent conversion coefficients. Neutron dose equivalents were also measured using a conventional BF3 detector based REM counter. Ambient dose equivalent measured by the REM counter is found to be in agreement with that determined from the neutron fluence spectra within their uncertainties. Angular distributions of the ambient dose equivalents were also determined from the measured fluence distributions at different angles.

Estimation of Ambient Dose Equivalent Rate Distribution Map Using Walking Survey Technique in Hirosaki City, Aomori, Japan.

At present, much emphasis is placed on the health risks associated with radioactivity present in the environment, especially since the accident at the Fukushima Daiichi Nuclear Power Plant. In this study, a walking survey was conducted in Hirosaki City using a NaI(Tl) scintillation spectrometer to estimate and map the distribution of the ambient dose equivalent rate to monitor the radiological safety of the general public in Hirosaki City, where many nuclear facilities are located nearby. The average (±standard deviation) ambient dose equivalent rate was 0.056 ± 0.020 µSv h-1. By comparison with the measurement data, it was found that the values of 85% of the data obtained using the walking survey technique deviated within ±20% relative to those obtained by spot measurements. Furthermore, the distribution of dose rates obtained in the nighttime survey was not significantly different from those obtained in the daytime.

Measurement of Ambient Dose Equivalent in Compact Proton Therapy using In-house Neutron Moderator-based Poly Allyl Diglycol Carbonate.

Purpose: The high-energy proton produces the unwanted dose contribution from the secondary neutron. The main purpose of this study is to report the validation results of in-house neutron moderator based on poly allyl diglycol carbonate (CR-39) detector, Chulalongkorn University Neutron Moderator (CUMOD) through the ambient dose equivalent, H*(10) measurement. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) Monte Carlo code was used to simulate the neutron response function. The CUMOD was calibrated with 241AmBe source calibrator in the range of 100-1000 µSv. The variation of neutron fields was generated employing different proton treatment plans covering most of the clinical scenarios. The ambient dose equivalents, H*(10), evaluated employing CUMOD were compared to those obtained with WENDI-II dosimeter. Results: The linear relationship between CUMOD and WENDI-II responses showed an R2 value close to 1. The H*(10) per Gy delivered dose was in the range of 22-105 µSv for a 10 cm × 10 cm field. Conclusion: The in-house CUMOD neutron moderator can expand the neutron detection dose range of CR-39 detector for ambient dose equivalent. The advantage of CUMODs is its capability to evaluate H*(10) in various positions simultaneously.

Benchmarking of stray neutron fields produced by synchrocyclotrons and synchrotrons in compact protontherapy centers (CPTC) using MCNP6 Monte Carlo code.

Proton therapy is an external radiotherapy using proton beams with energies between 70 and 230 MeV to treat some type of tumours with outstanding benefits, due to its energy transfer plot. There is a growing demand of facilities taking up small spaces and Compact Proton Therapy Centers (CPTC), with one or two treatment rooms, supposing the technical response of manufacturers to this request. A large amount of stray radiation is yielded in the interaction of proton beam used in therapy, neutrons mainly, hence, optimal design of shielding and verifications must be carried out in commissioning phases. Currently, almost 50 proton centers are under construction and start up in several countries, including ten in Spain. In the present work the effectiveness of shielding in two CPTC was verified with the Monte Carlo code MCNP6 by calculating the ambient dose equivalent, H*(10) due to secondary neutrons, outside the enclosures and walls of the center. The facilities modelled were the two centers currently operating in Spain, the first, since December 2019, with a superconductor synchrocyclotron, and the second, since March 2020, with a compact synchrotron. The geometry and materials are based on dimensions proposed a priori by the vendors, therefore, the paper is focused on check the suitability of the materials and thickness of the walls of the centers. Several models of the radiation sources were simulated, starting from a conservative assumptions, followed by more realistic scenarios. In all cases, the results reached for the ambient dose equivalent, H*(10), were below 1 mSv/year, which is the legal limit considered for the public in international references. Finally, considering that the recent ICRU Report 95 proposes changes in the operational quantities, the dose outside shieldingt has been evaluated in terms of the new next area surveillance quantity, H*, known as ambient dose, in the process of implementation.

A Geant4 shielding design for the first US carbon multi-ion hybrid synchrotron facility.

The Mayo Clinic Florida Integrated Oncology Building will be the home of the first spot-scanning only carbon/proton hybrid therapy system by Hitachi, Ltd. It will provide proton beams up to kinetic energies of 230 MeV and carbon beams up to 430 MeV n-1for clinical deployment. To provide adequate radiation protection, the Geant4 (v10.6) Monte Carlo toolkit was utilized to quantify the ambient dose equivalent at a 10 mm depth (H*(10)) for photons and neutrons. To perform accurate calculations of the ambient dose equivalent, three-dimensional computer-aided design files of the entire planned facility were imported into Geant4, as well as certain particle system components such as the bending magnets, fast Faraday cup, and gantry. Particle fluence was scored using 60 cm diameter spheres, which were strategically placed throughout areas of interests. Analytical calculations were performed as first-pass design checks. Major shielding slabs were optimized using Geant4 simulations iteratively, with more than 20 alternative designs evaluated within Geant4. The 430 MeV n-1carbon beams played the most significant role in concrete thickness Requirements. The primary wall thickness for the carbon fixed beam room is 4 meters. The presence of the proton gantry structure in the simulation caused the ambient dose equivalent to increase by around 67% at the maze entrance, but a decrease in the high energy beam transport corridor. All shielding primary and secondary goals for clinical operations were met per state regulation and national guidelines.

Application of liquid scintillation light guide (LSLG) to scattered X-ray measurement from dental panoramic radiography system.

A 1.2 m flexible liquid scintillation light guide (LSLG) detector connected to a portable electric device was developed and applied to monitor X-rays scattered from a panoramic dental X-ray imaging apparatus. The X-ray absorption of the LSLG tube was simulated for 20 and 40 keV X-rays. The LSLG detector was calibrated by comparing doses measured by a 3-inch NaI(Tl) detector using Am-241 and Cs-137 sources. A linear relationship was obtained between dose rates (µSv/h) and count rates (cps). The elapsed time profile of scattered radiation from a panoramic dental X-ray imaging apparatus was determined. Local absorbed doses were measured with small OSL dosimeters on the LSLG tube suspended over the shoulder and the eyeglass. Behind the ears and neck (in the center of the LSLG tube) showed high doses. The LSLG detector was useful for real-time monitoring of scattered X-rays.

Simulation of a new neutron calibration laboratory in Brazil using MCNP5.

Metrobras, a private company in Brazil, is developing a new laboratory for the calibration of neutron detectors in collaboration with the National Laboratory for Metrology of Ionizing Radiation (LNMRI). This work aimed at assessing neutron spectra, ambient dose equivalent rates, and neutron scattering in the calibration room of Metrobras Neutron Detectors Calibration Laboratory (LCDNM) using Monte Carlo simulation with MCNP5. Three models were simulated: vacuum, air, and the complete with the several materials employed in the actual room. Neutron spectra as a function of the distance from the source were obtained, compared, and used to calculate ambient dose equivalent at different points in the calibration room and results were compared to other neutron laboratories. Results show that LCDNM ambient dose equivalent rates are comparable to those of other neutron laboratories with 531.2 µSv . h-1 at 100 cm from the 592 GBq activity 241AmBe source. This work shows that LCDNM may be used as a calibration laboratory for neutron detectors in accordance with ISO 8529.

Neutron Radiation Dose Measurements in a Scanning Proton Therapy Room: Can Parents Remain Near Their Children During Treatment?

Purpose: This study aims to characterize the neutron radiation field inside a scanning proton therapy treatment room including the impact of different pediatric patient sizes. Materials and Methods: Working Group 9 of the European Radiation Dosimetry Group (EURADOS) has performed a comprehensive measurement campaign to measure neutron ambient dose equivalent, H*(10), at eight different positions around 1-, 5-, and 10-year-old pediatric anthropomorphic phantoms irradiated with a simulated brain tumor treatment. Several active detector systems were used. Results: The neutron dose mapping within the gantry room showed that H*(10) values significantly decreased with distance and angular deviation with respect to the beam axis. A maximum value of about 19.5 µSv/Gy was measured along the beam axis at 1 m from the isocenter for a 10-year-old pediatric phantom at 270° gantry angle. A minimum value of 0.1 µSv/Gy was measured at a distance of 2.25 m perpendicular to the beam axis for a 1-year-old pediatric phantom at 140° gantry angle.The H*(10) dependence on the size of the pediatric patient was observed. At 270° gantry position, the measured neutron H*(10) values for the 10-year-old pediatric phantom were up to 20% higher than those measured for the 5-year-old and up to 410% higher than for the 1-year-old phantom, respectively. Conclusions: Using active neutron detectors, secondary neutron mapping was performed to characterize the neutron field generated during proton therapy of pediatric patients. It is shown that the neutron ambient dose equivalent H*(10) significantly decreases with distance and angle with respect to the beam axis. It is reported that the total neutron exposure of a person staying at a position perpendicular to the beam axis at a distance greater than 2 m from the isocenter remains well below the dose limit of 1 mSv per year for the general public (recommended by the International Commission on Radiological Protection) during the entire treatment course with a target dose of up to 60 Gy. This comprehensive analysis is key for general neutron shielding issues, for example, the safe operation of anesthetic equipment. However, it also enables the evaluation of whether it is safe for parents to remain near their children during treatment to bring them comfort. Currently, radiation protection protocols prohibit the occupancy of the treatment room during beam delivery.

Spectro-dosemeter-based gamma dose rate network in Germany.

As a consequence of the Chernobyl accident in 1986 the Integrated Measurement and Information System (IMIS) was established (Weiss and Leeb, 1993) which includes on-line monitoring networks for the surveillance of radioactivity in Germany. Today, the German Federal Office for Radiation Protection (BfS) operates a gamma dose rate network with 1800 ambient dose equivalent rate H*(10) (ADER) stations almost equally distributed over the German territory. The ADER network integrates Geiger-Müller (GM) based detectors which, if low and high dose rate tubes are combined, are known to have excellent long-term stability and an extended dose rate range from environmental background level (20 nSv/h) up to several Sv/h. However, one main drawback is the lack of information about nuclides contributing to the observed dose rate. Therefore BfS has started to integrate LaBr3-based spectrometric detector systems (so-called spectro-dosemeters) in the existing ADER network. In this paper detector design, quality assurance and quality control (QA/QC) procedures are described as well as efforts required to characterize and operate monitoring networks based on spectrometric detectors.

Doses and spectra, of leaking-out and scattered photons from X-ray tubes used in dentistry.

Diagnosis by X-ray image are procedures widely used to improve the diagnosis or to follow the evolution of a medical procedure, also are used to support the cancer treatment with linear accelerators. The procedure involves the X-ray unit and the detection system, while the X-ray beam is projected onto the patient; along this procedure X-rays are scattered out from the patient body and X-rays leak-out from the X-ray unit. Both, the scattered and the leaking out radiation, are important during the evaluation or the design of the shielding barriers. X-ray tube features like voltage, electric current, target and filter are related to the dose due to the scattered and leakage radiation. Also, the dose due to the scattered radiation depends upon the scattering angle; while the dose due to the leaking out radiation, depend upon the X-ray tube housing. The objective of this work was to estimate, using Monte Carlo methods, the X-ray spectra and doses (air Kerma and Ambient dose equivalent) produced by scattered and leaking out radiation of 70, 80 and 90 kV X-ray units.

Effect of changing the neutron moderator in a thermal subcritical nuclear reactor.

Subcritical thermal nuclear reactor is the combination of nuclear fuel, moderator and external neutron source. Nevertheless, the amount of neutrons increases through nuclear fissions in the 235U in the fuel, the absorption and leakage of neutrons maintain the subcriticality condition. There are several subcritical reactors with different features and purposes. The open tank subcritical reactor Nuclear Chicago model 9000 is the heterogeneous combination of natural uranium, in hexagonal lattice, and light water, acting as moderator and reflector of neutrons, that uses a239PuBe neutron source. Worldwide there are several of these reactors mainly used in education, where due to the contact with water the Al tubes have corrosion. A possible solution is to use polyethylene as moderator. Aiming to evaluate the effect of changing the moderator, in this work Monte Carlo methods were used to model the reactor with polyethylene and water as moderators increasing the fuel load. For each model were estimated the keff and the µ parameters, and for Case 10a model was calculated the power, the Ambient dose equivalent and the fluence, of neutrons and γ-rays at five sites outside the reactor.

Development of a secondary standard ionization chamber for environmental level ambient dose equivalent measurement.

Ambient dose equivalent is the calibration quantity and type inspection quantity of environmental and workplace radiation monitoring instrument. For other unknown radiation fields other than the reference radiation field, the method of directly measuring H*(10) by secondary standard is particularly important. In order to ensure the reliability and accuracy of the value transfer of H*(10), an energy compensated high pressure ionization chamber for environmental level measurement was developed and used as a secondary standard. Firstly, the conversion coefficients from air kerma to ambient dose equivalent of X and γ-ray reference radiation field were calculated by Monte Carlo simulation. Then, the Monte Carlo model of ionization chamber was established, the thickness and energy compensation scheme of the outer wall of the chamber high voltage electrode was determined. Finally, the performance of the chamber were studied according to the calculated conversion coefficients. The results indicate that the relative deviation between calculated conversion coefficients and recommended value of ISO 4037-3 in the range of 15 keV-1.5 MeV is within ±4%. The ionization chamber meets the energy response requirements of ISO 4037-2 in the range of 65 keV-1.25 MeV. The linearity range of ambient dose equivalent rate is 0.8 µSv/h∼11.9 mSv/h, which can be used for a directly measurement of environmental level H*(10).

Dose calibration of Health Canada's Fixed Point Surveillance system for environmental radiation monitoring in terms of air kerma and H*(10).

The environmental radiation exposure in Canada has been monitored since 2002 by Health Canada's Fixed Point Surveillance network. The network consists of over eighty 7.6 cm × 7.6 cm sodium iodide spectrometers, and routinely reports to the public the environmental gamma radiation level throughout Canada. This paper describes the latest dose calibrations to air kerma and ambient dose equivalent for the future upgraded network. The calibration curves were developed using Monte Carlo techniques and further optimized via experiments in various reference fields. The dose calibration was validated over a wide range of gamma energy, dose measurement range, and angle of incidence under laboratory conditions. In environmental monitoring situations, the angular distribution of radiation exposure was analytically calculated by assuming a semi-infinite plume source, semi-infinite planar source, and infinite volume sources for the respective exposure scenarios of radioactive plume, ground contamination, and soil source. By coupling the resultant radiation angular distribution with detector's angular variation on dose response, the overall accuracy of dose measurement in each of these environmental scenarios was estimated. The accuracy is expected to be within ±3.7% for plume radiation, -5.6% for 137Cs ground contamination, and 0% to -17.1% for soil radioactive sources. The under-estimation for soil sources is mainly caused by absorption of radiation in the electronic system underneath the crystal.

Influence of solar activity on ambient dose equivalent H*(10) measured with thermoluminescent dosimeters in Slovenia.

Ambient dose equivalent H*(10) is measured to assess general population exposure to ionising radiation. From its spatial and time variations it is possible to identify sources of exposure. In Slovenia, semi-annual H*(10) is measured routinely with thermoluminescence dosimeters at 66 locations around the Nuclear Power Plant (NPP) Krsko and at 50 other locations covering the rest of Slovenian territory. Since the Chernobyl accident contamination had ceased to contribute to ambient dose equivalents, we have been calculating correlation coefficients between annual mean number of sunspots and annual H*(10). These correlation coefficients were calculated for five locations in western Slovenia and for five annual H*(10) extracted from measurements around NPP Krsko. Their ranges between -0.64 and -0.38 suggest a clear negative correlation between solar activity and H*(10). Mean annual H*(10) averted by solar activity in the past two solar maxima reached 0.070 mSv around NPP Krsko (155 m.a.s.l.) and 0.132 mSv and 0.180 mSv at Kredarica (2515 m.a.s.l.). Quantifying the influence of the solar activity on the ambient dose equivalent helps us to better understand exposure of the general population to ionising radiation.

Calculations for ambient dose equivalent rates in nine forests in eastern Japan from 134Cs and 137Cs radioactivity measurements.

Understanding the relationship between the distribution of radioactive 134Cs and 137Cs in forests and ambient dose equivalent rates (H˙∗(10)) in the air is important for researching forests in eastern Japan affected by the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident. This study used a large number of measurements from forest samples, including 134Cs and 137Cs radioactivity concentrations, densities and moisture contents, to perform Monte Carlo radiation transport simulations for H˙∗(10) between 2011 and 2017. Calculated H˙∗(10) at 0.1 and 1 m above the ground had mean residual errors of 19% and 16%, respectively, from measurements taken with handheld NaI(Tl) scintillator survey meters. Setting aside the contributions from natural background radiation, 134Cs and 137Cs in the organic layer and the top 5 cm of forest soil generally made the largest contributions to calculated H˙∗(10). The contributions from 134Cs and 137Cs in the forest canopy were calculated to be largest in the first two years following the accident. Uncertainties were evaluated in the simulation results due to the measurement uncertainties in the model inputs by assuming Gaussian measurement errors. The mean uncertainty (relative standard deviation) of the simulated H˙∗(10) at 1 m height was 11%. The main contributors to the total uncertainty in the simulation results were the accuracies to which the 134Cs and 137Cs radioactivities of the organic layer and top 5 cm of soil, and the vertical distribution of 134Cs and 137Cs within the 5 cm soil layers, were known. Radioactive cesium located in the top 5 cm of soil was the main contributor to H˙∗(10) at 1 m by 2016 or 2017 in the calculation results for all sites. Studies on the 137Cs distribution within forest soil will therefore help explain radiation levels henceforth in forests affected by the FDNPP accident. The merits of this study are that it modelled multiple forests for a long time period, with the important model inputs being informed by field measurements, and it quantified how the measurement uncertainties in these inputs affected the calculation results.