Dose Rate Map Research Articles

P121] Measurements of the scattered and leakage radiation in chest X-ray facilities

Purpose The chest X-ray (CXR) examination is the most frequently employed radiographic procedure for screening and diagnosis of diseases because it provides substantial clinical information at a low cost and minimal radiation exposure. Despite its enormous benefits, the ionizing nature of x-rays poses potential health risks to patients, medical personnel and general public. In this study, safety assessment of CXR facilities of selected public hospitals was performed through scattered and leakage radiation measurements. Methods Scattered radiation measurements of CXR machines were conducted using a survey meter. Dose rates were obtained from 20.7 cm polymethylmethacrylate (PMMA) phantom for both postero-anterior (PA) and antero-posterior (AP) projections. Dose rate maps were generated at distances of 50, 100, 150 and 200 cm at different angles around the phantom. Measurements performed in the control room using typical clinical irradiation conditions were used to obtain the calculated annual doses of the facilities. Leakage radiation measurements were also performed at 100 cm distance from the focus of the four sides of the X-ray tube with the collimators completely closed. Measured and calculated dose rates were evaluated with respect to the acceptable limits set by the International Commission on Radiological Protection (ICRP). Results For both CXR projections, the scattered radiation dose decreased as the distance from the center of the phantom was increased. The largest drop was observed at an angle of 225° with the dose rate at 200 cm dropping to 1.1% of the dose rate in 50 cm. Moreover, the AP projection produced higher scattered radiation than the PA projection due to the shorter source-to-image distance (i.e., 180 cm for PA and 100 cm for AP). The highest leakage radiation dose rate was 5.969 μ Sv/hr and the calculated annual dose rates were 1846 nSv/yr and 20.34 nSv/yr. These values were found to be below the leakage radiation and occupational dose limits set by ICRP which are 1000 μ Sv/hr and 20 × 10 6 nSv/yr, respectively. Conclusions Through the scattered and leakage radiation measurements, it was proven that existing procedures in the facilities evaluated do not pose significant threats on patients, medical personnel and the general public.

Assessment of Radiation Dose to Workers Resulting from External Exposure to Potassium in NORM Industries in Korea

Potassium is a naturally-occurring radioactive material (NORM). Potassium contains 40K, a natural radionuclide that may enhance the radiation dose to workers in potassium processing facilities. The objective of the present study is to assess radiation dose to workers resulting from external exposure to potassium in NORM industries in Korea. Five potassium processing facilities were selected for this dose assessment study. Initial dose rate mapping was made through the whole facility site for scanning purposes. Based on an analysis of the initial dose rate mapping and potassium processing, the areas requiring more detailed dose rate measurements were selected. Target workers for radiation dose assessment were selected based on the second step dose rate mapping results and specific working conditions. The annual radiation dose to the target workers was assessed using the measured dose rates and hypothetical exposure scenarios. Annual radiation dose to the workers resulting from external exposure to potassium ranged between 0.003 - 0.364 mSv/yr. The results of the present study can be used to better protect workers handling naturally-occurring radioactive materials.

Virtual Reality: Lessons learned from WEST design and perspectives for nuclear environment

From 2012–2016, the Tore Supra tokamak was upgraded in an x-point divertor device within the frame of the WEST project. A huge design activity was made to modify the whole configuration of Tore Supra including the in-vessel components. Some of the most noticeable changes are the addition of inner divertor coils inside the vacuum vessel and the replacement of all plasma facing components.Fitting all the new components in a previously existing environment was probably one of the major challenges of the project. To improve the assembly plan and the design of toolings, Virtual Reality (VR) was introduced in the design process.For complex assembly operations, simulations were developed for VR headsets from CAD models. From the requirements expressed by the design team, the VR engineers developed dedicated tools to address the specific issues. Engineers, designers and operators were then able to experience the assembly conditions in a 3D 360° real size immersive environment and interact with the models thanks to a large set of dedicated tools which are developed in the VR lab. This approach showed impressive results for issues such as accessibility, complex kinematics of large components and design of handling tools. It allowed reducing the number of mock-ups and helped the WEST assembly team to reduce the number of unforeseen assembly issues. It was for example used to simulate the assembly of the divertor: insertion through the ports, toroidal displacement on rails, and assembly with gaps lower than 1 mm.Virtual Reality will also provide innovative solutions to help designers in the anticipation of challenges linked to operation of a nuclear facility: operator’s training for nuclear maintenance, virtual mock-ups for qualification of maintenance, maintenance scenario optimization with real time simulations… A first module plotting the dose rate is currently being tested. On the one hand, it would for example allow designers to improve the design by adapting maintenance operations to any complex 3D map of dose rate, and on the other hand, it would permit workers to train for operations in radioactive environments (showing no go areas, hot spots…).

Assessment of radiation dose received by nuclear plant personnel through a video-based surveillance system

This work aims at developing a video-based surveillance system for safety purposes in nuclear plants. The objective is to assess the radiation dose received by nuclear plant personnel, while they execute daily tasks, by means of computer vision methods. The system is conceived to provide some redundancy to the radioprotection means currently in use, being independent and complementary to them. After evaluating some methods from the literature for automatic target detection and tracking, a novel system is developed to correctly detect, track and identify people, so that the radiation dose received by each person is reliably computed. The video data are supplied by cameras installed in the nuclear plant room. Radiation dose rate mapping is combined with the tracking results to account for the received doses. We provide experimental results from a research reactor room, which show that the proposed system achieves radiation dose estimates that are in general similar to the ones of the ground truth. The database developed in this work for performance evaluation has been made publicly available for the research community.

Technical Note: DoseMapper: A validated GUI-based exact numerical modeling method of shielding in PET/CT facilities.

To provide a faster and more intuitive way of designing shielding for PET facilities, while still relying on the principles of the AAPM 108 Taskforce guidelines, as well as illustrating the calculation output using dose maps that are easily evaluated. A graphical user interface was developed, implementing an inverse AAPM method, wherein radiation sources and shield barriers are manually defined. Simulations are calculated using a user-defined control mesh grid. DoseMapper simulations were verified against manual calculations using the AAPM guidelines, as well as compared with in situ dose rate measurements using four different dosemeters. DoseMapper simulations were virtually identical to manual calculations using AAPM guidelines, with a maximum relative error of <0.01%. Comparison with in situ measurements showed that DoseMapper-simulated dose rates in all instances are higher than what can be measured, ensuring that no unintended hotspots can be overlooked in the shielding design. DoseMapper is an easy to use implementation of the AAPM 108 Taskforce principles that allows for a rapid iterative design process of shielding in PET facilities, and the resulting maps of dose rate and annual accumulated dose serve as clear documentation for the design.

Dose Rate Mapping by Small UAV to Radioactive Contamination Area

Dose Rate Mapping by Small UAV to Radioactive Contamination Area

Representativeness of individual external doses estimated for one quarter of residents in the Fukushima Prefecture after the nuclear disaster: the Fukushima Health Management Survey

After the Fukushima Dai-ichi Nuclear Power Plant accident, the Fukushima Health Management Survey (FHMS) was launched. The Basic Survey, a component of FHMS, is a questionnaire used to survey residents across the Fukushima Prefecture about their behaviour in the first 4 months after the accident. The questionnaire findings are used to determine individual external doses by linking behaviour data to a computer programme with daily gamma ray dose rate maps, drawn after the accident. Through 30 June 2015, the response rate was only 27.2% (558 550 population), indicating that the findings might not be generalisable because of poor representativeness of the population. The objective of this study was to clarify if the data from the FHMS Basic Survey were representative of the entire population, by conducting a new survey to compare the external doses between non-respondents and respondents in the previous survey. A total of 5350 subjects were randomly selected from 7 local regions of Fukushima Prefecture. An interview survey was conducted with the non-respondents to the FHMS Basic Survey. A total of 990 responses were obtained from the previous non-responders by interview survey. For the regions Kempoku, Kenchu, Kennan, Aizu, Minami-Aizu, Soso, and Iwaki, differences in mean effective dose (95% confidence interval) in mSv between the non-responders and previous responders were 0.12 (0.01–0.23), −0.09 (−0.21–0.03), −0.06 (−0.18–0.07), 0.05 (−0.04–0.14), 0.01 (−0.01–0.02), 0.09 (0.01–0.17), 0.09 (0.00–0.17), respectively. The differences fall neither within the interval (−∞, −0.25) nor within the interval (0.25, ∞). These findings imply that mean effective doses between the previous and new respondents were not different, with a significantly indifferent region of 0.25 mSv according to equivalence tests. The present study indicates that the dose distribution obtained from about one-quarter of Fukushima residents represents the dose distribution for the entire Fukushima Prefecture.

Radioprotection shielding for neutrons induced by the reaction (2H (40 MeV), 12C)

In the framework of design studies for SPIRAL2, the simulation of the neutron flux generated by 40 MeV deuterons on a thick 12 C target was performed and compared to experimental data. The calculation of the dose rate of these neutrons allowed to compare four materials being considered for radioprotection shielding: barites, gypsum, ordinary concrete and heavy concrete. The simulated map of the neutron dose rate in the production building shows a very high dose rate around the neutron source and in the environment of some of the accelerator equipment.

Terrestrial gamma radiation dose rate in Cienfuegos, Cuba

This study assesses the level of background radiation for Cienfuegos Province, Cuba. Measurements of outdoor gamma radiation (of terrestrial and cosmic origin) in air were performed at 198 locations using a GPS navigator and a dose meter (SRP-68-01, 30 × 25 mm NaI detector). The average absorbed dose was found to be 73.9 nGy h -1 (17.2–293.9 nGy h -1 ), corresponding to an annual effective dose of 74.7 μ Sv (21–324 μ Sv). When compared with the data available for other places, the absorbed gamma doses obtained in this study indicate a background radiation level that falls within natural limits for the Damuji, Salado and Caonao watersheds; however, the Arimao and Gavilanes watersheds present levels of the absorbed dose and annual effective dose comparable with high background radiation areas. An isodose map of the terrestrial gamma dose rate in Cienfuegos was drawn using the GIS application “Arc View”. This study provides important baseline data of radiation exposure in the area.

Dose rate mapping of VMAT treatments

Human tissues exhibit a varying response to radiation dose depending on the dose rate and fractionation scheme used. Dose rate effects have been reported for different radiations, and tissue types. The literature indicates that there is not a significant difference in response for low-LET radiation when using dose rates between 1 Gy min−1 and 12 Gy min−1 but lower dose rates have an observable sparing effect on tissues and a differential effect between tissues. In intensity-modulated radiotherapy such as volumetric modulated arc therapy (VMAT) the dose can be delivered with a wide range of dose rates.In this work we developed a method based on time-resolved Monte Carlo simulations to quantify the dose rate frequency distribution for clinical VMAT treatments for three cancer sites, head and neck, lung, and pelvis within both planning target volumes (PTV) and normal tissues.The results show a wide range of dose rates are used to deliver dose in VMAT and up to 75% of the PTV can have its dose delivered with dose rates <1 Gy min−1. Pelvic plans on average have a lower mean dose rate within the PTV than lung or head and neck plans but a comparable mean dose rate within the organs at risk. Two VMAT plans that fulfil the same dose objectives and constraints may be delivered with different dose rate distributions, particularly when comparing single arcs to multiple arc plans.It is concluded that for dynamic plans, the dose rate range used varies to a larger degree than previously assumed. The effect of the dose rate range in VMAT on clinical outcome is unknown.

The SATRAM Timepix spacecraft payload in open space on board the Proba-V satellite for wide range radiation monitoring in LEO orbit

The Space Application of Timepix based Radiation Monitor (SATRAM) is a spacecraft platform radiation monitor on board the Proba-V satellite launched in an 820km altitude low Earth orbit in 2013. The is a technology demonstration payload is based on the Timepix chip equipped with a 300µm silicon sensor with signal threshold of 8keV/pixel to low-energy X-rays and all charged particles including minimum ionizing particles. For X-rays the energy working range is 10–30keV. Event count rates can be up to 106cnt/(cm2s) for detailed event-by-event analysis or over 1011cnt/(cm2s) for particle-counting only measurements. The single quantum sensitivity (zero-dark current noise level) combined with per-pixel spectrometry and micro-scale pattern recognition analysis of single particle tracks enables the composition (particle type) and spectral characterization (energy loss) of mixed radiation fields to be determined. Timepix׳s pixel granularity and particle tracking capability also provides directional sensitivity for energetic charged particles. The payload detector response operates in wide dynamic range in terms of absorbed dose starting from single particle doses in the pGy level, particle count rate up to 106–10/cm2/s and particle energy loss (threshold at 150eV/µm). The flight model in orbit was successfully commissioned in 2013 and has been sampling the space radiation field in the satellite environment along its orbit at a rate of several frames per minute of varying exposure time. This article describes the design and operation of SATRAM together with an overview of the response and resolving power to the mixed radiation field including summary of the principal data products (dose rate, equivalent dose rate, particle-type count rate). The preliminary evaluation of response of the embedded Timepix detector to space radiation in the satellite environment is presented together with first results in the form of a detailed visualization of the mixed radiation field at the position of the payload and resulting spatial- and time-correlated radiation maps of cumulative dose rate along the satellite orbit.

Radioactive Doses — Predicted and Actual — and Likely Health Effects

Five years have passed since the nuclear accident at Fukushima Daiichi Nuclear Power Stations on 11 March 2011. Here we refer to reports from international organisations as sources of predicted values obtained from environmental monitoring and dose estimation models, and reports from various institutes in Japan are used as sources of individual actual values. The World Health Organization, based on information available up to 11 September 2011 (and published in 2012), reported that characteristic effective doses in the first year after the accident, to all age groups, were estimated to be in the 10–50 mSv dose band in example locations in evacuation areas. Estimated characteristic thyroid doses to infants in Namie Town were within the 100–200 mSv dose band. A report from the United Nations Scientific Committee on the Effects of Atomic Radiation published in 2014 shows that the effective dose received by adults in evacuation areas during the first year after the accident was 1.1–13 mSv. The absorbed dose to the thyroid in evacuated settlements was 7.2–35 mSv in adults and 15–83 mSv in 1-year-old infants. Individual external radiation exposure in the initial 4 months after the accident, estimated by superimposing individual behaviour data on to a daily dose rate map, was less than 3 mSv in 93.9% of residents (maximum 15 mSv) in evacuation areas. Actual individual thyroid equivalent doses were less than 15 mSv in 98.8% of children (maximum 25 mSv) in evacuation areas. When uncertainty exists in dose estimation models, it may be sensible to err on the side of caution, and final estimated doses are often much greater than actual radiation doses. However, overestimation of the dose at the time of an accident has a great influence on the psychology of residents. More than 100 000 residents have not returned to the evacuation areas 5 years after the Fukushima accident because of the social and mental effects during the initial period of the disaster. Estimates of radiation doses placed in the public domain must be based on scientific evidence, and the way such information is communicated to residents should be carefully considered to avoid psychosocial effects that may have a greater bearing on health than the radiation itself.

Radiation survey on Fukushima Medical University premises about four years after the Fukushima nuclear disaster.

This study was conducted on the Fukushima Medical University (FMU) premises (in Fukushima City, Fukushima Prefecture) about four years after the Fukushima Daiichi Nuclear Power Plant accident. Its objectives were (1) to create a map of the ambient gamma dose rate (air-kerma rate) distribution, (2) to evaluate the air-kerma rate originating from natural radionuclides, and (3) to investigate the effects of snow cover on changes in the air-kerma rate. This man-borne survey revealed that the air-kerma rate varies widely, ranging from 0.038 μGy h(－1) to 0.520 μGy h(－1), and is higher on grass than on the other investigated surface types, such as soil, asphalt, and bricks. In this area, the mean air-kerma rate from natural radiation was evaluated to be 0.03 ± 0.01 μGy h(－1), which is close to 0.04 μGy h(－1), which was measured in central Fukushima City by a local authority.Furthermore, snowfall was found to reduce the air-kerma rate by 5%-30%. This reduction was attributed to attenuation of the primary radiation while passing through the snow cover, and the measured contribution of scattered radiation to the air-kerma rate reduction was small. The reduction rate was found to depend on the initial snow depth but to maintain a similar value for a couple of days, after the snow had partially melted and its depth had decreased. Finally, analysis of the daily dose due to external exposure received on the FMU premises revealed that no further health effects due to chronic radiation exposure at this site are to be expected.

Dose rate estimates and spatial interpolation maps of outdoor gamma dose rate with geostatistical methods; A case study from Artvin, Turkey

In this study, compliance of geostatistical estimation methods is compared to ensure investigation and imaging natural Fon radiation using the minimum number of data. Artvin province, which has a quite hilly terrain and wide variety of soil and located in the north–east of Turkey, is selected as the study area. Outdoor gamma dose rate (OGDR), which is an important determinant of environmental radioactivity level, is measured in 204 stations. Spatial structure of OGDR is determined by anisotropic, isotropic and residual variograms. Ordinary kriging (OK) and universal kriging (UK) interpolation estimations were calculated with the help of model parameters obtained from these variograms. In OK, although calculations are made based on positions of points where samples are taken, in the UK technique, general soil groups and altitude values directly affecting OGDR are included in the calculations. When two methods are evaluated based on their performances, it has been determined that UK model (r = 0.88, p < 0.001) gives quite better results than OK model (r = 0.64, p < 0.001). In addition, as a result of the maps created at the end of the study, it was illustrated that local changes are better reflected by UK method compared to OK method and its error variance is found to be lower.

MCR2S unstructured mesh capabilities for use in shutdown dose rate analysis

As nuclear fusion progresses towards a sustainable energy source and the power of tokamak devices increases, a greater understanding of the radiation fields will be required. As well as on-load radiation fields, off-load or shutdown radiation field are an important consideration for the safety and economic viability of a commercial fusion reactor. Previously codes such as MCR2S have been written in order to predict the shutdown dose rates within, and in regions surrounding, a fusion reactor. MCR2S utilises a constructive solid geometry (CSG) model and a superimposed structured mesh to calculate 3-D maps of the shutdown dose rate. A new approach to MCR2S calculations is proposed and implemented using a single unstructured mesh to replace both the CSG model and the superimposed structured mesh.This new MCR2S approach has been demonstrated on three models of increasing complexity. These models were: a sphere, the ITER computational shutdown dose rate benchmark and the DEMO computational shutdown dose rate benchmark. In each case the results were compared to MCR2S calculations performed using MCR2S with CSG geometry and a superimposed structured mesh. It was concluded that the results from the unstructured mesh implementation of MCR2S compared well to the CSG structured mesh calculations. It was found that the resolution of the unstructured mesh can significantly affect the results of the calculations, and therefore it is important to finely mesh components with streaming paths or significantly contribute to the dose rate in the areas of interest. Although computationally more expensive it was found that there are some clear advantages when using unstructured meshes with MCR2S calculations which are outlined in this paper.

Assessment of fusion facility dose rate map using mesh adaptivity enhancements of hybrid Monte Carlo/deterministic techniques

Three mesh adaptivity algorithms were developed to facilitate and expedite the use of the CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques in accurate full-scale neutronics simulations of fusion energy systems with immense sizes and complicated geometries. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility and resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation. Additionally, because of the significant increase in the efficiency of FW-CADIS simulations, the three algorithms enabled this difficult calculation to be accurately solved on a regular computer cluster, eliminating the need for a world-class super computer.

Evaluation of external radiation exposure of workers engaged in loading of radioactive material shipping cask between ship and truck by crane

In the marine transportation of radioactive material shipping casks, a loading/unloading cargo work has been carried out using a large crane between a ship and a truck at the interface of land and sea. At that time there is a radiation exposure situation where a worker should approach a cask for handling, tie-down and dosimetric inspection. For a spent fuel cask whose surface dose rate is relatively high, it is necessary to evaluate a radiological risk for cargo workers as one of the planned exposure situations since there is no valid shielding between cask and worker for a long time. In order to instantly evaluate worker dose under any working environment, it is necessary to prepare a simple method that accurately calculate a dose rate map around a cask of interest. In this study, simple calculation methods developed for a spent fuel cask have been surveyed to determine the best fit parameters for a representative type-B wet cask of Japan NFT14P. Then, worker dose is practically evaluated by using the parameters under working environment of cask loading in Japan. There assumed to be three types of cargo workers; handler, inspector and spotter. It is found that the highest individual dose is evaluated to 1.70 μSv per cask loading for an inspector on a carrier of truck and the annual cumulative dose will be well below the annual public limit of 1mSv. The collective dose per loading is amounted to about 17 man-μSv in a total of 19 workers. The occurrence of latent effects can be negligible but the dose values are still higher compared to estimated exposures for escort and off-link population along a road route in U.S. land transportation of spent fuel.

A high spatial resolution outdoor dose rate map of the Rio de Janeiro city, Brasil, risk assessment and urbanization effects

Geophysical surveys have been widely used for geological mapping, and with appropriate ground calibration the database can be converted to radiometric surveys. Herein we present a case study of a high resolution map of the City of Rio de Janeiro, Brazil, constructed by integrating aerial and in situ (ground) measurements. This allowed assessing the influence of urbanization observed between 1978 and 2009 on radiation external exposure. Radiometric, geological and demographics databases were integrated in a Geographic Information System (GIS) facility. The average absorbed dose rate recorded for the city was 96 ± 0.3 nGy h−1 and ranged from 22 to 152 nGy h−1 among districts. The corresponding annual effective dose was determined as 0.09 ± 0.01 mSv. The number of people at risk per year according to the estimated natural radiation levels was 17 ± 1.4, among the 28,000 new cases estimated by the INCA (Brazilian National Cancer Institute) within a population of approximately 6.3 million people.

Analysis of Decay Dose Rates and Dose Management in the National Ignition Facility

A detailed model of the Target Bay (TB) at the National Ignition Facility (NIF) has been developed to estimate the post-shot radiation environment inside the facility. The model includes the large number of structures and diagnostic instruments present inside the TB. These structures and instruments are activated by neutrons generated during a shot, and the resultant gamma dose rates are estimated at various decay times following the shot. A set of computational tools was developed to help in estimating potential radiation exposure to TB workers. The results presented in this paper describe the expected radiation environment inside the TB following a low-yield DT shot of 10(16) neutrons. General environment dose rates drop below 30 μSv h(-1) within 3 h following a shot, with higher dose rates observed in the vicinity (~30 cm) of few components. The dose rates drop by more than a factor of two at 1 d following the shot. Dose rate maps of the different TB levels were generated to aid in estimating worker stay-out times following a shot before entry is permitted into the TB. Primary components, including the Target Chamber and diagnostic and beam line components, are constructed of aluminum. Near-term TB accessibility is driven by the decay of the aluminum activation product, 24Na. Worker dose is managed using electronic dosimeters (EDs) self-issued at kiosks using commercial dose management software. The software programs the ED dose and dose rate alarms based on the Radiological Work Permit (RWP) and tracks dose by individual, task, and work group.

NIRS external dose estimation system for Fukushima residents after the Fukushima Dai-ichi NPP accident

The great east Japan earthquake and subsequent tsunamis caused Fukushima Dai-ichi Nuclear Power Plant (NPP) accident. National Institute of Radiological Sciences (NIRS) developed the external dose estimation system for Fukushima residents. The system is being used in the Fukushima health management survey. The doses can be obtained by superimposing the behavior data of the residents on the dose rate maps. For grasping the doses, 18 evacuation patterns of the residents were assumed by considering the actual evacuation information before using the survey data. The doses of the residents from the deliberate evacuation area were relatively higher than those from the area within 20 km radius. The estimated doses varied from around 1 to 6 mSv for the residents evacuated from the representative places in the deliberate evacuation area. The maximum dose in 18 evacuation patterns was estimated to be 19 mSv.