Measurements of environmental terrestrial gamma radiation average dose rate in three mountainous locations in the western region of Saudi Arabia

This paper describes measurements of external gamma radiation average dose rate from terrestrial gamma rays 1 m above the ground in three different mountainous locations in the western region of the Kingdom of Saudi Arabia. These locations are, At-Taif city, Al-Hada village and Ash-Shafa village. CaSO(4):Dy (TLD-900) thermoluminescence dosemeters were used for the detection of terrestrial gamma radiation at 40 different places in the three locations. The values of terrestrial gamma radiation average dose rate measured ranged between 14 and 279 nG h(-1) for the time interval from June 2001 to June 2002. The measured average dose rate varied with the season of the year. The gamma radiation average dose rates were 468, 541 and 781 microGy y(-1) for At-Taif city, Al-Hada village and Ash-Shafa village, respectively. The corresponding average effective dose rates for the three locations were 94, 108 and 156 microSv y(-1), respectively.     

Measurement of natural radioactivity and dose rate assessment of terrestrial gamma radiation in the soil of southern Punjab, Pakistan

Activity concentrations of (226)Ra, (232)Th, (40)K and (137)Cs in soil samples collected from the most populous area of southern Punjab of Pakistan have been measured by gamma-ray spectrometry. The measured activity concentrations for these radionuclides are compared with the reported data from different other countries and it is found that measured activity concentrations are comparable with the worldwide measured average values reported by the UNSCEAR. Subsequently assessed radiological effects show that the mean radium equivalent activity (Ra(eq)) is 96.7 +/- 15.2 Bq kg(-1) and air absorbed dose rate (D) is 46.1 +/- 7.3 nGy h(-1). The values of internal and external radiation hazard indices are found to be less than unity. The annual effective radiation dose is calculated to be 0.28 +/- 0.05 mSv, which is well below the limit of 1.0 mSv y(-1) recommended by the International Commission on Radiological Protection, for the general public.     

Evaluation of absorbed dose rate and annual effective dose equivalent due to terrestrial gamma radiation in rocks in a part of Southwestern Nigeria

The average outdoor absorbed dose rate in air and the average annual effective dose equivalent due to terrestrial gamma radiation from 40K, 238U and 232Th in rocks in Ondo and Ekiti States, Southwestern Nigeria have been evaluated from measurements of the concentrations of these radionuclides in this environmental material. The concentration measurements were obtained using a very sensitive gamma spectroscopic system consisting of a 7.6 cm x 7.6 cm NaI(Tl) scintillation detector coupled to a computerised ACCUSPEC installation. The average absorbed dose rate and average annual effective dose equivalent was found to be 8.33 +/- 2.76 nGy x h(-1) and 8.7 +/- 2.9 microSv x y(-1) respectively.     

Background gamma terrestrial dose rate in Nigerian functional coal mines

Measurements of the background terrestrial gamma radiation dose rates at different indoor and outdoor locations on the surfaces of Okpara underground and Okaba open cast mines in Nigeria were made. Two duly calibrated low-level gamma survey metres were held 1 m above the ground surface for these measurements. Measurements were also made at various locations inside the mine tunnel at the Okpara mine. Results indicate that the indoor background gamma radiation is comparable for both mining environments. The mean outdoor gamma dose rate determined for the Okaba mining environment is 10.4 nGy h(-1) as against 11.7 nGy h(-1) for Okpara. The ranges are 8.5-16.5 nGy h(-1) for the Okpara measurements and 7.5-14.0 nGy h(-1) for Okaba. Thus, the outdoor gamma dose rates appear to be generally lower at the Okaba open cast mine than at Okpara. The indoor dose rate values range from 11.0 to 17.0 nGy h(-1) in both environments. These indoor measurements have nearly the same mean values 14.4 and 14.5 nGy h(-1) for Okpara and Okaba environments, respectively. The indoor to outdoor dose rate ratio is 1.2 for Okpara and 1.4 for Okaba. These values are in consonance with the corresponding ratio given in literature. Dose rate measurements inside the mine tunnel at the Okpara mine are higher than the surface indoor measurements ranging from 13.5 to 20.5 nGy h(-1) with a mean of 16.5 nGy h(-1). The higher dose rate values measured in the mine tunnel are attributable to the concentration of radon in the 'closed' environment of the mine tunnel.     

Gis predictive mapping of terrestrial gamma radiation in the Northern state, Sudan

This study presents the evaluation of absorbed dose in air due to gamma-emitting nuclides from (238)U and (232)Th series, (40)K and (137)Cs and the corresponding geographical information system (GIS) predictive mapping for the Northern State. Activity concentration of (238)U, (232)Th , (40)K and (137)Cs in soil samples collected from different locations have been measured using high-resolution gamma spectrometry. On average, activity concentrations were 19±4 ((238)U), 47±11 ((232)Th), 317±65 ((40)K) and 2.26 Bq kg(-1) for (137)Cs. Absorbed dose rate in air at a height of 1 m above ground surface was calculated using seven sets of dose rate conversion factors (DRCFs) and the corresponding annual effective dose was estimated. On average, the values obtained fall within a narrow range of 44 and 53 nGy h(-1), indicating that the variation in absorbed dose rate is insignificant for different DRCFs. The corresponding annual effective dose ranged from 53 to 65 μSv y(-1). Using GIS, prediction maps for concentrations of (238)U, (232)Th, (40)K and (137)Cs were produced. Also, a map for absorbed dose rate in air at a height of 1 m above the ground level was produced, which showed a trend of increasing from the west towards south-east of the State.     

Assessment of terrestrial gamma radiation doses for some Egyptian granite samples

External exposures of population to ionising radiation due to naturally occurring radionuclides in sixty-three granite samples from three different locations in south eastern desert of Egypt were considered in this article. Average outdoor gamma dose rates in air were 190, 290 and 330 nGy h(-1) for Elba, Qash Amir and Hamra Dome granites, respectively. The corresponding doses in indoor air are 270, 400 and 470 nGy h(-1), respectively. These average values give rise to annual effective dose (outdoor, indoor and in total) 0.24, 1.4 and 1.6 mSv for Elba granite. For Qash Amir and Hamra Dome granites the corresponding values were 0.35, 2 and 2.3 mSv and 0.41, 2.3 and 2.7 mSv, respectively.     

Dose rate dependency of electronic personal dosemeters measuring X- and gamma-ray radiation

Three models of electronic personal dosemeters (EPDs)-Siemens Mk 2.3, Rados RAD-60S and Vertec Bleeper Sv-were irradiated with seven photon beam qualities: 60Co, 137Cs and the ISO narrow spectrum series X-ray qualities N-250, N-200, N-150, N-60 and N-20. The personal dose equivalent rates delivered to the devices varied between 0.002 and 0.25 mSv s(-1). Measurements were made with the EPDs mounted free-in-air as well as against Lucite and water phantoms. Results for all models of EPDs showed differences in personal dose equivalent energy response for different energies covered by this range of radiation qualities, with different models showing variations from 15 to 65%. In some cases, the personal dose equivalent rate response of these devices varied by a factor of 3 between irradiations at typical calibration dose rates and those normally encountered by nuclear energy workers.     

Extended survey of indoor and outdoor terrestrial gamma radiation in Greek urban areas by in situ gamma spectrometry with a portable Ge detector

The results obtained from more than 1000 indoor and outdoor in situ gamma spectrometry measurements in 41 towns (from all geographic subdivisions) of the Greek mainland (not islands) are presented. From the in situ gamma spectra the absorbed dose rate in air due to uranium series, thorium series, 40K and 137Cs are derived and discussed.     

Occupational radiation protection dosimetry in Nigeria

The general features of occupational radiation protection dosimetry in Nigeria within the period 1990-1999 have been summarised. About 640 personnel, representing about 25% of the estimated number of radiation workers in Nigeria, were monitored by the TL dosimetry technique during the period, with the majority being the personnel of the teaching hospitals across the country. Most private establishments, especially the X ray diagnostic centres, operate without dosimetry coverage or supervision by a regulatory authority. The weighted mean of the annual effective dose ranged between 0 and 28.97 mSv with the upper limit of collective effective dose being 18.47 man.Sv per year. The individual risk estimate due to this is about 1.5 x 10(-3) per year and this was among the medical personnel. The value could be more if all radiation workers in the country were monitored.     

Dose dependences of radiation induced yield in mixed radiation fields

A theoretical model that describes dose dependences of trap filling (radiation yield) in mixed radiation fields consisting of two components is proposed. The model consists of one type of electron traps and one type of hole traps and assumes as an initial step the creation of two types of tracks, each represented by some volume with a uniform electron-hole pair density, different for each track. The relaxation process that follows comprises interband recombination, trapping of electrons and holes, and recombination of electrons with trapped holes and of holes with trapped electrons. These processes result in filled traps in amounts depending on the absorbed dose in the track and the number and types of tracks created in a given region of irradiated matter. The summation over the matter with areas of different degrees of overlapping (assuming poisson distribution of the created tracks), gives expressions for the dependences of trap filling as a function of doses for separated and simultaneous irradiation. It is shown that the key parameters determining the behaviour of the dose dependences are the ratios between the doses in the separated tracks and the average doses delivered on the irradiated matter by the separated components of the mixed field. If the ratios of the average dose to the track dose are low, the dose dependences will be linear. In the opposite limiting case the dose dependencies go to saturation. The linear and additive approximations of dose dependences in a mixed field are valid at low doses only.     

Aspects of EIT with gamma radiation

Abstract

We have observed the effect of electromagnetically induced transparency (EIT) in a nuclear system. A series of Mössbauer measurements was performed on a single crystal of natural FeCO₃. When the condition of nuclear level crossing is fulfilled, a clear reduction in absorption or transparency is seen at the point of level crossing. This reduced absorption cannot be ascribed to saturation effects because of the specific geometry of the experiment. We present an explanation in terms of level mixing by a small non-axial component of the electric field gradient. It can be shown that such a scheme is equivalent, at least in principle, with the schemes considered for EIT in quantum optics.     

Dose quantities in radiation protection and dosemeter calibration

Dosimetric concepts and the definition of dose quantities for use in radiation protection were defined by the ICRP and the ICRU. Three types of quantities are of relevance for radiation protection purposes: basic physical quantities, protection quantities, and operational quantities. The physical quantities are universally accepted for the characterisation of radiation fields. These are defined in any point of the field, and the units are directly obtained by primary standards. The protection quantities form the basis for dose limitation. These quantities are not directly measurable. A set of measurable operational quantities provides, in general, a conservative estimate of the protection quantities. These operational quantities are recommended in routine monitoring of occupational exposure. The relation between these different types of quantities is discussed. In addition, the calibration procedure for dosemeters in terms of the operational quantities, including the use of calibration phantoms, is summarised.     

Dose quantities in radiation protection and their limitations

For more than 50 years the quantity absorbed dose has been the basic physical quantity in the medical applications of ionising radiation as well as radiological protection against harm from ionising radiation. In radiotherapy relatively high doses are applied (to a part of the human body) within a short period and the absorbed dose is mainly correlated with deterministic effects such as cell killing and tissue damage. In contrast, in radiological protection one is dealing with low doses and low dose rates and long-term stochastic effects in tissue such as cancer induction. The dose quantity (absorbed dose) is considered to be correlated with the probability of cancer incidence and thus risk induced by exposure. ICRP has developed specific dosimetric quantities for radiological protection that allow the extent of exposure to ionising radiation from whole and partial body external radiation as well as from intakes of radionuclides to be taken into account by one quantity. Moreover, radiological protection quantities are designed to provide a correlation with risk of radiation induced cancer. In addition, operational dose quantities have been defined for use in measurements of external radiation exposure and practical applications. The paper describes the concept and considerations underlying the actual system of dose quantities, and discusses the advantage as well as the limitations of applicability of such a system. For example, absorbed dose is a non-stochastic quantity defined at any point in matter. All dose quantities in use are based on an averaging procedure. Stochastic effects and microscopic biological and energy deposition structures are not considered in the definition. Absorbed dose is correlated to the initial very short phase of the radiation interaction with tissue while the radiation induced biological reactions of the tissue may last for minutes or hours or even longer. There are many parameters other than absorbed dose that influence the process of cancer induction, which may influence the consideration of cells and/or tissues at risk which are most important for radiological protection.     

Assessment of gamma dose rates from terrestrial exposure in Serbia and Montenegro

The gamma dose rates due to naturally occuring terrestrial radionuclides ((226)Ra, (232)Th and (40)K) were calculated based on their activities in soil samples, determined by gamma-ray spectrometry. A total of 140 soil samples from 21 different regions of Serbia and Montenegro were collected. The gamma dose rates ranged from 7.40 to 29.7 nGy h(-1) for (226)Ra, from 12.9 to 46.5 nGy h(-1) for (232)Th and from 12.5 to 37.1 nGy h(-1) for (40)K. The total absorbed gamma dose rate due to these radionuclides varied from 34.5 to 97.6 nGy h(-1) with mean of 66.8 nGy h(-1). Assuming a 20% occupancy factor, the corresponding annual effective dose varied from 4.23 x 10(-5) to 11.9 x 10(-5) Sv with mean of 8.19 x 10(-5) Sv, i.e. the dose was lower than world wide average value. According to the values of external hazard index (mean: 0.39) obtained in this study, the radiation hazard was found to be insignificant for population living in the investigated area.     

Evaluation of indoor gamma radiation dose in dwellings

The use of materials containing naturally occurring radionuclides for house construction may enhance the natural radiation background to which some population groups are exposed. External exposure results from gamma emitter radionuclides existing in the walls, floor and ceiling. Mathematical models can be used to predict external dose rates inside a room, provided the compartment geometry and the radionuclide concentration activities are known. This paper presents a methodology and a computer code for theoretical evaluation of indoor external gamma doses in the air. The room was modelled as three pairs of rectangular slabs of finite thickness. Doses were evaluated by applying a photon transport model, taking into account self-absorption and radiation build-up. Calculations were performed for 40K, 226Ra and 232Th, considering concrete walls. The results obtained show good agreement with those reported in the literature. Dose conversion factors are presented in a practical manner, ready to use for radiological impact screening.     

Dose assessment of aircraft crew in The Netherlands

As the operator of the National Dose Registration and Information System, NRG has implemented a system for radiation exposure monitoring for the Dutch airlines. The system is based on the use of computer generated flight plans together with dose calculations using the CARI-6M program. Before installing the system a study was performed to estimate the uncertainty in the assessment of the annual dose of the crew members. It was concluded that the proposed system complies with international recommendations on the uncertainty in dose assessments in individual monitoring and that the operational costs of the system are low.     

Dose build up correction for radiation monitors in high-energy bremsstrahlung photon radiation fields

Conventional radiation monitors have been found to underestimate the personal dose equivalent in the high-energy bremsstrahlung photon radiation fields encountered near electron storage rings. Depth-dose measurements in a water phantom were carried out with a radiation survey meter in the bremsstrahlung photon radiation fields from a 450 MeV electron storage ring to find out the magnitude of the underestimation. Dose equivalent indicated by the survey meter was found to build up with increase in thickness of water placed in front of the meter up to certain depth and then reduce with further increase in thickness. A dose equivalent build up factor was estimated from the measurements. An absorbed dose build up factor in a water phantom was also estimated from calculations performed using the Monte Carlo codes, EGS-4 and EGSnrc. The calculations are found to be in very good agreement with the measurements. The studies indicate inadequacy of commercially available radiation monitors for radiation monitoring within shielded enclosures and in streaming high-energy photon radiation fields from electron storage rings, and the need for proper correction for use in such radiation fields.     

Analysis of radiation workers' dose records in the Korean National Dose Registry

Data are presented on the externally received personal dose equivalent for radiation workers who used the Korea Radioisotope Association's personal monitoring and dose record keeping service since 1984, and provide initial statistics on Korean workers who have been exposed to ionising radiation in different occupations. The total number of workers registered during the period of 1984 to 1999 was 64,518. The number increased steadily and the accumulated dose also increased. The proportion of radiation workers by occupation was 38.4% for nuclear power plants, 20.3% for industrial organisations and 12.4% for non-destructive industry. The annual collective dose of radiation workers was 31.72 man.Sv in 1999. The mean annual dose by sex was 1.49 mSv for males and 0.56 mSv for females and the mean annual dose for a worker was 1.41 mSv with the highest mean dose being received by non-destructive industry (3.53 mSv). Very few workers (0.8%) received more than 20 mSv (2 rem) and only one more than 50 mSv, the legal limit for an annual dose increase. There has been a steady decline in the mean dose since 1984, showing a significant decrease in dose with time (p<0.001). The data showed that radiation protection in Korea was improving, though annual doses were still higher than other countries.