Total Effective Dose Rates Research Articles

Activity concentration of NORM at reclaimed ex-coal mines in South Sumatra, Indonesia

Unbeknownst to us, natural radiation is in close proximity to human existence. Post-mining practices can generate naturally occurring radionuclides that have the potential to cause radiological effects on human health. This study utilized soil samples obtained from ex-mining reclamation land in Banko Barat, South Sumatra Province, to assess the concentration of natural radioactivity and the corresponding radiation dose. Eight composite soil samples were collected from reclaimed land. The samples underwent processing and analysis through gamma spectrometry in order to ascertain the activity concentration of the radionuclides 226Ra, 232Th, and 40K. The RESRAD-ONSITE 7.2 software utilizes the activity concentration of soil samples as an input parameter to accurately predict the dose rate for workers. The projected dose rate was determined by considering four routes of exposure: external gamma radiation, inhalation, radon, and ingestion of soil. The measured activity values of soil samples from reclamation land for 226Ra ranged from 25.6704 ± 0.5082 to 58.8555 ± 0.6917 Bq/kg. The activity values for 232Th ranged from 30.5546 ± 0.1757 to 42.53020.3004 Bq/kg. Lastly, the activity values for 40K ranged from 533.902 ± 6.4023 to 988.081 ± 9.6948 Bq/kg. The total effective dose rate measured in the reclaimed land falls between 0.08673 and 0.1469 millisieverts per year. The value does not surpass the maximum limit set by BAPETEN, which is one mSv/year. Therefore, no additional protective measures are required for workers.

Evaluation of radiological health risk caused by the use of fly ash in cement and concrete production and its storage

ABSTRACT As a result of firing pulverized coal in thermal power plants, enormous amounts of fly ash (FA) are produced as industrial waste. The release into the atmosphere and storage of this industrial waste remains one of the major environmental problems that threaten human health by contributing to air, water, and soil pollution. The recovery and reuse of FA in the construction industry is the only economic solution to the existing problem. In this study, the potential radiological risk caused by the usage of FA in concrete and cement production as a main component and its storage in landfill sites was evaluated for people and works by estimating radiological parameters (activity concentration and alpha index, annual effective doses, and the corresponding excess lifetime cancer risks) based on activity concentrations of terrestrial radionuclides in FA. Also, the radiological risk to the workers working in the FA landfill site was evaluated using the Residual Radioactivity Onsite 7.2 code. The average activity concentrations of terrestrial radionuclides in FA samples from the Tunçbilek lignite coal-fired thermal power plant at Kütahya province of Turkey were measured as 417, 156 and 454 Bq kg−1 for 226Ra, 232Th and 40K, respectively. When using up to 35% by mass of FA in cement and concrete, the average values of the radiological parameters revealed that they were within the recommended safety limits. However, code estimations showed that a regular worker in FA storage would be exposed to a total effective dose rate greater than 3 mSv y−1.

Second Target Station Bunker Shielding with Populated and Unpopulated Neutron Beamlines: Preliminary Design

At the U.S. Department of Energy’s Oak Ridge National Laboratory, the Second Target Station (STS) beamline sources for preliminary design have been used to perform a shielding analysis of the bunker. Prompt total effective dose rates (i.e., neutron plus photon effective dose rates when the proton beam is on) were calculated on top of the bunker roof and outside the bunker wall. These areas outside the bunker will be generally accessible, so the prompt total dose rate in these areas should not exceed 2.5 μSv‧h−1 (0.25 mrem‧h−1). This paper presents the required shielding thicknesses to meet this dose rate limit. In one instance, this dose rate limit is not met: For a combination of populated and unpopulated beamlines, the prompt total dose rate outside the bunker across from the unpopulated beamline, which has less shielding because of the lack of beamline shielding, slightly exceeds 2.5 μSv‧h−1. Once more details are known regarding the STS high-density concrete density and composition, a future analysis will investigate the shielding modifications required to reduce the calculated prompt total dose rates for this configuration to less than 2.5 μSv‧h−1.

Verification of the NACAC atmospheric dispersion calculation using a hypothetical accident in a neighboring nuclear power plant

Calculating atmospheric dispersion in a hypothetical nuclear accident plays an essential role as a predictive tool in nuclear emergency planning. The Nuclear Accident Consequence Analysis Code (NACAC) has been developed as an in-house code by the Thailand Institute of Nuclear Technology to develop its capability for understanding the transboundary radiological effect of a hypothetical nuclear accident. To verify the performance of the NACAC, a comparison between the calculated results using the NACAC with those from the well-developed Java-based Real-time On-line Decision Support (JRODOS) code is performed with representative cases of all seasons in a year. Two hypothetical severe accidents at the Fangchenggang pressurized water reactor, namely, (i) a loss of off-site power and (ii) a large-break loss of coolant, are utilized as the initial boundary conditions. The air concentration characteristic maps obtained from both codes are analyzed, revealing that the dominant radionuclide dispersion pathways are similar in all study cases. The air concentration, ground concentration, and total effective dose rate at several locations up to 500 km away from the release point are directly compared. Typically, the calculated results from the two codes are in reasonable agreement. However, for specific locations and directions, the NACAC predictions of the concentrations and total effective dose rate are lower than those from the JRODOS. These differences may be influenced by the different trajectory calculations between the two codes, which resulted in varying plume densities in certain areas. However, it does not significantly affect the radiation effect assessment. The NACAC still provides a total effective dose rate corresponding to the JRODOS, lower than one mSv throughout the main dispersion direction.

Estimation of granite radiation hazards of Deh Siahan village in Rafsanjan city

There are villages in Rafsanjan city (Kerman, Iran) that are nearby or located on radioactive granite rock, unfortunately. In this study, by measuring the 232Th, 238U and 40K radioactive elements, the amount of radiation hazards caused by radioactive granite rock. For the residents Deh Siahan village has been investigated. The results show that the radiation hazard parameters of Deh Siahan village are higher than the permissible values announced by the UNSCEAR reports. For example, the total effective dose rates of Deh Siahan village was determined to be 0.30 ± 0.01 mSv/y. The calculated values of the excess lifetime cancer risks (ELCR) for Deh Siahan village is 10.51 ± 0.03× 10-4. The average absorbed dose rate in air for Deh Siahan village is 156.68 ± 4.01 nGy/h for the whole studied samples, this value is about 285% of the maximum level recommended by (ICRP60, 1991) to the public (1 mSv/y), so there is high radiological risk for the peoples in those village.

Simulating the Potential Impacts of Nuclear Power Plant Accident for Northern Vietnam

The Fangchenggang nuclear power plant has been built very close to the Vietnam boundary. This is done to generate potential impacts for Northern Vietnam if nuclear power plant accident occurs. This study applied the Weather Research and Forecasting (WRF) model to construct the meteorological data at horizontal mesh resolution of 1 km as input for the FLEXible PARTicle dispersion model (FLEXPART). The assumption of the nuclear accident at Fangchenggang Power Plant is considered with setup parameter of the Fukushima accident. The results show a similar in simulating the 137Cs concentration from 03 out of 24 experiments configured with different parameterisation schemes of the WRF model. However, the dry and wet deposition of radioactive 137Cs are significantly different. It is especially illustrated that if the accident occurs, then almost all provinces in northern Vietnam are affected. The high concentration of radioactive pollutants may be intensively transported from Fangchenggang nuclear power plant to Vietnam under the domination of wind fields in the wintertime. The maximum values of the total effective dose rate could reach up to over 10 mSvh-1 of dose rate during 50 to 100 hours. Importantly, the maximum effective dose continues to be observed during 145 to 205 hours.

Natural and fallout radioactivity mapping of Kakrapar Gujarat site, India

This study presents the activity and outdoor gamma absorbed dose rates (terrestrial and cosmic) due to the naturally occurring (226Ra, 232Th, and 40K) and anthropogenic (137Cs) radionuclides around Kakrapar Gujarat site. The activity level (Bq/kg) in the soil ranged 5.7–48.4 for 226Ra, 9.7–28.0 for 232Th, 83.9–585.7 for 40K, and 0.2–4.6 for 137Cs, respectively. The mean concentration levels measured in Kakrapar soil from the naturally occurring radioisotopes are lower than the corresponding global average values. 137Cs activity in the soil is comparable with the preoperational period. The total effective dose rates in air outdoors ranged 17.2–78.1 μSv/y with the mean value of 32.8 μSv/y. The mean value of absorbed dose rate due to cosmic components was 37.5 nGy/h which is comparable to the worldwide reported values.

Gamma-ray measurements of the activated target components in a cyclotron used for positron emission tomography

The proton irradiation in a PET cyclotron produces radioactive by-products with high levels of activity in the target components, which are strongly dependent on the specific parameters used in the cyclotron facility. Because the target assembly parts must be replaced periodically, external exposure to operators must be assessed. In this work, high-resolution gamma-ray spectrometry was used to determine the activity levels for the specific radionuclides induced in the target components of a medical 9.6 MeV MINItrace cyclotron. The focus was on three target components, which are periodically replaced: Havar foil, Titanium foil and the Helicoflex seal located on the path of the proton beam just in front of the silver vessel containing the enriched water. Eight radionuclides were identified in the Havar foil: 54Mn, 56Co, 57Co, 58Co, 60Co, 95Tc, 109Cd and 183Re. In the titanium foil, the nuclides 46Sc, 48V, 51Cr, 56Co, 57Co, and 75Se were identified, while only 51Cr, 58Co, and 60Co were detected in the Helicoflex seal. The total effective dose rates to which staff are exposed at the moment of replacing the target were calculated. The waste management for these target components was also considered, using the half-lives of the detected radionuclides in order to determine the time necessary for the activity levels in the target components to be below the exemption levels recommended by the IAEA.

Risk assessment of atmospheric and liquid discharges of radionuclides on humans, marine biota and terrestrial wildlife

Egypt is currently considering building a nuclear power plant (NPP). Consequently, an environmental impact assessment is required for potential releases to the environment. The aim of this work is to implement an environmental impact assessment for a nuclear power plant to assess radiation doses to man and the wildlife, to aid the decision-making related to the environment effects of that facility. This study focused on atmospheric dispersion modeling code cap88-Pc, ERICA Tool version 1.2 and R&D128 approach for predicting the activity concentrations of released radionuclides and the corresponding dose rates to human, as well as terrestrial and marine wildlife at different distances from the reactor at normal operation. The results of total dose rate to marine wildlife, recommended that the release point to be beyond 1600 m offshore for such organisms to be protected and the summing risk quotient to be well below unity. The risk quotients for terrestrial organisms from deposited radionuclides on the ground and the emitted noble gases were well below unity. The total effective dose rate to man from air emissions and from consumption of contaminated marine species (fish and crustaceans) from liquid discharges combined was 5.19E−3 mSv y−1 corresponding to 2.6E−7 cancer risk factor, which is acceptably low risk. The radionuclide Co-60 was one of the major contributors of dose rate to marine organisms (beside Co-58 and Mn-54) and of total effective dose for human together with Ag-110m and C-14. The current study considered to provide the baseline information for further implementation of regulatory policies when site specific data is available.

Survey of Activity Concentration and Dose Estimation of Naturally Occurring Radionuclides (232Th, 238U and40K) in the Coastal area of Dakar, Senegal

Objectives: The present study has been carried out to assess the activities concentration of some terrestrial gamma emitting radionuclides and to evaluate the radiation radiological hazard. Methods/Statistical Analysis: In-situ gamma ray spectrometer with a large high-density Bismuth Germanate crystal detector was used for field measurement. Under assumption of secular equilibrium of 238Th and 238U activity concentration and associated radiological hazard indices such as Radium equivalent activity (Raeq) and External Radiation hazards (Hex) and gamma dose rate were estimated. Findings: The following means (±standard error) of concentration (BqKg-1) were obtained: 260.57±31.17 for 40K, 12.82±0.75 for 232Th and 17.04±2.21 for 238U. The Raeq values ranging from 26.10 BqKg-1 to 117.61 BqKg-1 ; the Hex ranged 0.07 Bqkg1 to 0.31 BqKg-1; the total effective dose rates varied from 12.55 nGy.h-1 56.15 nGy.h-1 while the indoor and the outdoor maximum value was 0.27 mSv and 0.07 mSv, respectively. This study is the first evaluation of reference background levels of terrestrial gamma emitting radionuclides in this area. Application/Improvements: Measured primordial radionuclides are essential data for radiological assessment and radioprotection studies as well as for nuclear safety regulatory bodies. Keywords: Gamma Dose Rate, Naturally Occurring Radionuclides, Radiological Hazard

Measurement of natural radioactivity in granites and its quartz-bearing gold at El-Fawakhir area (Central Eastern Desert), Egypt

The distribution of natural radionuclides (226Ra, 232Th and 40K) in Granites and its quartz-bearing gold at El-Fawakhir area (Central Eastern Desert, Egypt) were measured by using γ-ray spectroscopy [NaI (Tl) 3″ × 3″]. X-Ray Fluorescence technique was used for chemical analyses of the studied samples. The specific activity of 226Ra, 232Th and 40K values are in range (3 ± 0.5 to 43 ± 2 Bqkg−1), (5 ± 0.7 to 41 ± 2 Bqkg−1) and (128 ± 6 to 682 ± 35 Bqkg−1) respectively. The absorbed dose rates ranged from 13.8 to 58.4 nGy h−1, where the total effective dose rates were determined to be between 16.7 and 70.9 μSvy−1. The maximum external hazard index (Hex) is 0.3 nGyh−1. The calculated values of the excess lifetime cancer risks (ELCR) and annual effective dose rate values are in between (8.48 × 10−5 and 2.63 × 10−4) and (24.2 and 72.9 μSvy−1) respectively. Geochemically, the studied granites consist of major oxides, they are characterized by SiO2, K2O, Na2O, Al2O3, and depleted in CaO, MgO, TiO2, and P2O5. The average absorbed dose rate (Do) in air is 37.8 nGyh−1 for the whole studied samples, this value is about 3.78% of the 1.0 mSvy−1 recommended by (ICRP-60,1991) to the public, so there is no radiological risk for the workers in that area.

Assessment of Radiological Dose around a 3-MW TRIGA Mark-II Research Reactor

A hypothetical accidental case of a 3-MW TRIGA Mark-II research reactor has been assumed to assess the radiological consequences due to the deposition of 137Cs and 90Sr on ground, vegetation, milk and meat. The air concentrations in sixteen cardinal directions have been estimated where the maximum concentration has been found to be at 110 m distance from the core of the reactor for all the directions. Calculated maximum doses of 137Cs, 90Sr and both 137Cs and 90Sr have been found to be within the ranges of 0.005-0.014 μSv hr–1, 0.013-0.036 μSv hr–1 and 0.018-0.05 μSv hr–1, respectively for all the directions, which are below the measured background dose limit 0.25 μSv hr–1 and also within the IAEA acceptable dose rate limit of 0.5 μSv hr–1. The calculated low doses due to the aforementioned radionuclides can be considered negligible with regard to the radiation hazards. The relationship between total effective dose rate for various pathways (i.e. immersion, inhalation, ground deposition, and ingestion of contaminated vegetation, milk, meat) and air concentration in all the directions has been established. Obtained relation reveals that the total effective dose rate is directly proportional to the air concentration, and the overall proportionality constants for 137Cs and 90Sr radionuclides have been obtained as 0.57 and 0.28, respectively. This study might provide information on the radiological safety required for the radiation protection purposes of the people living in the vicinity of the reactor site.

Assessment of Radiological Dose around a 3-MW TRIGA Mark-II Research Reactor

A hypothetical accidental case of a 3-MW TRIGA Mark-II research reactor has been assumed to assess the radiological consequences due to the deposition of 137Cs and 90Sr on ground, vegetation, milk and meat. The air concentrations in sixteen cardinal directions have been estimated where the maximum concentration has been found to be at 110 m distance from the core of the reactor for all the directions. Calculated maximum doses of 137Cs, 90Sr and both 137Cs and 90Sr have been found to be within the ranges of 0.005-0.014 μSv hr–1, 0.013-0.036 μSv hr–1 and 0.018-0.05 μSv hr–1, respectively for all the directions, which are below the measured background dose limit 0.25 μSv hr–1 and also within the IAEA acceptable dose rate limit of 0.5 μSv hr–1. The calculated low doses due to the aforementioned radionuclides can be considered negligible with regard to the radiation hazards. The relationship between total effective dose rate for various pathways (i.e. immersion, inhalation, ground deposition, and ingestion of contaminated vegetation, milk, meat) and air concentration in all the directions has been established. Obtained relation reveals that the total effective dose rate is directly proportional to the air concentration, and the overall proportionality constants for 137Cs and 90Sr radionuclides have been obtained as 0.57 and 0.28, respectively. This study might provide information on the radiological safety required for the radiation protection purposes of the people living in the vicinity of the reactor site.

Chemical and radiological characterization of fly and bottom ash landfill of the former sulfate pulp factory Plaški and its surroundings

The subject of this study was chemical and radiological characterization of the fly and bottom ash, by-product of the combustion of coal used as an energy source in the former sulfate pulp factory in Plaški. The research involves determination of the concentration of macro, micro and trace elements and activities of the radionuclides in: (i) ash from different positions of the landfill; (ii) soil samples in the zone of the influence of the landfill; (iii) control soil samples and (iv) sediment sample from the river Dretulja. Besides, in situ measurement of an effective dose rate above ash/soil was also determined. In relation with the control soil the average increase of the concentrations of the elements Ca, Cd, Hg, Ni, Se, Sr, Th and U in the samples taken from the fly and bottom ash landfill as well as soil samples within the radius of 300 m from the landfill was 38.3, 6.7, 9.9, 8.5, 9.4, 7.2, 3.6 and 5.7 times, respectively. In these samples, the concentrations of the above mentioned elements were in the following ranges: calcium from 7.94 to 19.7 %; cadmium from 0.33 to 1.66 mg/kg; mercury from 0.18 to 0.49 mg/kg; nickel from 260 to 1500 mg/kg; selenium from 2.7 to 21 mg/kg; strontium from 176 to 542 mg/kg; thorium from 8 to 55 mg/kg and uranium from 5.6 to 19.7 mg/kg. Compared to the world's average soil concentration, uranium and thorium values increased 3.7 and 1.7 times, respectively. The mean value of the total effective dose rate measured in the air at the height of 1 m for all samples of ash and soil under the influence of the landfill was 1.60 mSv/yr. Compared to the Croatian average (0.7015 mSv/yr), the determined mean value for the Plaški landfill is two times higher. However, compared to the local background (0.14 mSv/yr), the mean value of the total effective dose rate measured above the Plaški landfill is 11.4 times higher. In the samples of ash and contaminated soil regardless of the sampling location the activity concentrations of the radionuclides in Bq/kg vary in the following ranges: 226Ra from 82.10 to 314.90 (mean value 145.99), 232Th from 32.50 to 223.60 (mean value 76.76) and 238U from 69.10 to 243.20 (mean value 134.38). Compared to the mean values found in the background soil 226Ra and 238U mean activity concentrations increased from 1.6 to 6.4 times and 232Th from 1.4 to 4.3 times. In order to reduce total effective dose rate to the local “background” values and to prevent redistribution of the radionuclides and heavy metals from the deposited material into the environment fly and bottom ash landfill must be sealed with 10 cm thick layer of the material with low permeability.

Gamma radiation measurements of naturally occurring radioactive samples from commercial Egyptian granites

Due to the widespread use of granites as building and ornamental materials, measurements of natural radioactivity for a total 27 selected samples of commercial granites used in Egypt were carried out by using a high pure germanium detector. The activity concentrations of 226Ra, 232Th and 40K of commercial granites ranged from 25 to 356, 5 to 161, and 100 to 1,796 (Bq kg−1), respectively. The concentrations of these radionuclides are compared with the international recommended values. To evaluate the radiological hazard of the natural radioactivity, the radium equivalent activity, the absorbed dose rate, the effective dose rate, and the hazard index have been calculated. The radium equivalent activity Raeq varied from 41 to 669 (Bq kg−1) which exceeds the permitted value (370 Bq kg−1) and the internal hazard index H in varied from 0.2 to 2.8 which is higher than 1. The absorbed dose rate due to the natural radioactivity in the samples under investigation ranged from 19 to 310 (nGy h−1). The total effective dose rates per person indoors were determined to be 0.09 to 1.5 (mSv year−1).

Study of 7Be activity concentration in the air of northern Spain

Time series analysis techniques have been applied to the study of the 7Be activity concentration in air, which was obtained by weekly collecting aerosols in filters in the north of Spain, and measuring them by gamma ray spectrometry. Between 2001 and 2009, a trend pattern is found. The cyclical component of this series is clearly influenced by the solar cycle (11 years period). Two kinds of seasonality were found in the time series: annual periodicity and variations with seasons of the year. The irregular component has also been studied but its contribution has been insignificant; one of its possible causes, rainfall, has been analysed and the conclusion is that the atmospheric concentration of 7Be is unaffected by rainfall. The contribution of 7Be presence in air to the total effective dose rate in air has also been studied, resulting negligible.

中間貯蔵容器を配置したモジュラー型遮蔽庫の合理的放射線遮蔽性能評価

This paper discusses a rational methodology to evaluate dose rate at the boundary of the interim storage facility using two methods; 1) a method taking into account “shade effect” which represents the self-shielding effect of interim storage containers, 2) utilization of a newly developed simplified code. It was proved that the method of the shade effect is applicable to the secondary gamma-ray, which dominates approximately 50% of the total effective dose rate. Thus, the effective dose rate attributed to the secondary gamma-ray depends on both the number of containers in the facility and the shade coefficient which represents the factor of the shade effect. Using this shade coefficient, the dose rate was estimated for the facility that stored 8 or 16 containers around the boundary of the facility and the results were consistent with that of the Monte Carlo calculation. In the present study, a simple neutron transport code, MCNP-ANISN_W, was newly developed aiming at the simplification of large-scale calculation. The results of the dose rate agreed well with that of the Monte Carlo calculation. The results of the present study show that the simple evaluation technique and the code developed in this study would be useful for the evaluation of dose rate around the boundary of the interim storage facility.

Natural radioactivity distribution in geological matrices around Kaiga environment

The activity and absorbed dose rate of the naturally occurring radionuclides, viz. 238U, 232 Th and 40K were determined in soil and rock samples collected around Kaiga site. The mean activity levels (Kaiga soil) of naturally occurring 232 Th are comparable with that in worldwide soil, while concentrations of 238U and 40K are lower than those in worldwide soil. The absorbed dose rate in outdoor air ranged 20-58 nGy . h-1 with a mean of 33.3 nGy . h-1, which is below the world average of 60 nGy . h-1. The total effective dose rate in outdoor air for soils ranged 25.6-74.4 mSv . y-1 with a mean of 43.0 mSv . y-1. The estimated dose rate at Kaiga is comparable with that estimated at Kakrapar and Rawatbhata and much less than that estimated at coastal sites of India.

Natural radiation doses for cosmic and terrestrial components in Costa Rica

A study of external natural radiation, cosmic and terrestrial components, was carried out with in situ measurements using NaI scintillation counters while driving along the roads in Costa Rica for the period July 2003–July 2005. The geographical distribution of the terrestrial air-absorbed dose rates and the total effective dose rates (including cosmic) are represented on contour maps. Information on the population density of the country permitted the calculation of the per capita doses. The average effective dose for the total cosmic component was 46.88±18.06 nSv h −1 and the average air-absorbed dose for the terrestrial component was 29.52±14.46 nGy h −1. The average total effective dose rate (cosmic plus terrestrial components) was 0.60±0.18 mSv per year. The effective dose rate per capita was found to be 83.97 nSv h −1 which gives an annual dose of 0.74 mSv. Assuming the world average for the internal radiation component, the natural radiation dose for Costa Rica will be 2.29 mSv annually.

Determination of natural radioactivity in building materials used in Tunisian dwellings by gamma ray spectrometry

The radioisotopic content of 17 samples of natural and manufactured building materials collected in Tunisia have been analysed by using gamma spectrometry. From the measured gamma ray spectra, activity concentrations are determined for (232)Th, (226)Ra, (235)U and (40)K. The total effective dose and the activity concentration index are calculated applying the dose criteria recommended by the European Union for building materials. The results of (226)Ra, (232)Th and (40)K found in Tunisian building materials indicate that radium and thorium concentrations do not exceed 40 Bq kg(-1), but potassium concentration varies between 50 and 1215 Bq kg(-1). The total effective dose rates per person indoors are determined to be between 0.07 and 0.86 mSv y(-1). Only two materials exceed the reference level of 0.3 mSv y(-1). The activity concentration index is <1.