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During a routine whole body counting measurement of a worker at the Nuclear Research Center Negev, abnormal activities of (232)Th and (238)U were measured. After a thorough investigation, it was found that the radioactivity was due to a rubber bracelet ('balance bracelet') worn by the worker during the measurement. The bracelet was counted directly by an high pure germanium gamma spectrometry system, and the specific activities determined were 10.80 ± 1.37 Bq g(-1) for (232)Th and 5.68 ± 0.88 Bq g(-1) for natural uranium. These values are obviously high compared with normally occurring radioactive material (NORM) average values. The dose rate to the wrist surface was estimated to be ∼3.9 µGy h(-1) and ∼34 mGy for a whole year. The dose rate at the centre of the wrist was estimated to be ∼2.4 µGy h(-1) and ∼21 mGy for a whole year. The present findings stresses a more general issue, as synthetic rubber and silicone products are common and widely used, but their radioactivity content is mostly uncontrolled, thus causing unjustified exposure due to enhanced NORM radioactivity levels.

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Cherenkov counting of the 9°Sr/9°Y pure beta emitters is an attractive method for 9°Sr activity determination, but the color quenching effect may be significant, especially for strongly colored or semi-opaque media. A quench correction method based on the external source of some liquid scintillation systems (named ESAR - external source area ratio) was proposed and checked for aqueous solutions and was proved to be effective also for urine samples. In the present work, the application of the ESAR method for fast determination of 9°Sr/9°Y activity in milk samples is described.

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Cherenkov counting of the (90)Sr-(90)Y pure beta emitters in aqueous samples is an attractive method; but color quenching correction is needed, this being especially significant for urine which is characterized by a strong coloration. A quench correction method based on the external source of some liquid scintillation systems (named ESAR-External Source Area Ratio) was proposed for aqueous solutions. In the present work, the application of the ESAR method for determination of (90)Sr-(90)Y in human urine samples is described.

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The color quenching correction for Cerenkov counting is normally based on indexes defined for liquid scintillation spectra. Because of the differences in the spectrum shapes, these indexes are not optimal, and a different method was proposed based on the integral counting of the external (152)Eu source spectra of a 1200 Quantulus system. We have shown that this method, named ESAR (external source area ratio), is superior to all the other correction methods, offering the highest sensitivity and the widest range of application.

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The Cherenkov counting efficiency varies strongly with color quenching, thus correction curves must be used to obtain correct results. The external (152)Eu source of a Quantulus 1220 liquid scintillation counting (LSC) system was used to obtain a quench indicative parameter based on spectra area ratio. A color quench correction curve for aqueous samples containing (90)Sr/(90)Y was prepared. The main advantage of this method over the common spectra indicators is its usefulness also for low level Cherenkov counting.

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Inhalation is the main route of internal exposure to radioactive aerosols in the nuclear industry. To assess the radiation dose from the intake of these aerosols, it is necessary to know their physical (aerodynamic diameter distribution) and chemical (dissolution rate in extracellular lung fluid) characteristics. Air samples were taken from the uranium processing plant at the Nuclear Research Center, Negev. Measurements of aerodynamic diameter distribution using a cascade impactor indicated an average activity median aerodynamic diameter value close to 5 microm, in accordance with the recent recommended values of International Commission on Radiological Protection (ICRP) model. Solubility profiles of these aerosols were determined by performing in vitro solubility tests over 100 d in a simultant solution of the extracellular fluid. The tests indicated that the uranium aerosols should be assigned to an absorption between Types M and S (as defined by the ICRP Publication 66 model).

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For dose assessment of workers at Nuclear Research Center Negev exposed to natural uranium, spot urine samples are analysed and the results are normalised to 24-h urine excretion based on 'standard' man urine volume of 1.6 l d(-1). In the present work, the urine volume, uranium level and creatinine concentration were determined in two or three 24-h urine collections from 133 male workers (319 samples) and 33 female workers (88 samples). Three volunteers provided urine spot samples from each voiding during a 24-h period and a good correlation was found between the relative level of creatinine and uranium in spot samples collected from the same individual. The results show that normalisation of uranium concentration to creatinine in a spot sample represents the 24-h content of uranium better than normalisation to the standard volume and may be used to reduce the uncertainty of dose assessment based on spot samples.

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Sampling of radioiodine in air is accomplished by passing an air sample through an activated charcoal cassette. A mathematical model was developed, which assumes that radioiodine distribution along the cassette axis can be expressed by an exponential function. The model was validated experimentally for cassettes used for air sampling of radioiodine production boxes. There is a good agreement between the measurements and model predictions. Furthermore, when breakthrough occurs, the model can be used to estimate the activity that passed through the cassette unabsorbed.

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The Organic Bound Tritium (OBT) amount in the body may induce changes in the evaluation of the internal exposure to tritium, due to its different retention time relative to HTO. OBT measurements for urine are not performed routinely, mainly because of the lengthy work needed in preparation of the samples, when using the standard oxygen combustion method. A simpler and more rapid method based on liquid scintillation counting (LSC) was employed, and an evaluation was performed to check its suitability for urine samples. The principle of the proposed method is based on subtracting the LSC counts of the water phase sample (HTO) from the total activity of the urine sample (OBT+HTO). A good correlation was found between the results obtained when applying the proposed direct method and the Gold standard method. The minimum detectable activity when using this method was determined.

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A practical and simple method was employed to improve the minimum detectable activity (MDA) for lung counting measurements by summing several accumulated spectra. The method was checked for natural uranium, which produces peaks due to photon energies of 63.3, 92.6 and 185.7 keV. By combining nine measurements, an overall improvement of the MDA by a factor of about 3 was achieved. Uranium contamination levels lower than the MDA of a single spectrum could be detected with acceptable accuracy when analyzing the sum spectra. Specific results are given for four workers occupationally exposed to natural uranium.

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Considerable errors in the determination of radioactivity in lungs are caused by the usual assumption of homogeneous distribution, while actually the radioactive contamination can be distributed, and even in the form of a 'hot spot'. Modern lung counter systems use several HPGe detectors, and the ratio of the different count rates of the detectors can be used to locate a 'hot spot' and apply correction algorithms for activity calculation. In the present work it is shown that the variability of the isotopic composition of the uranium source has a major effect on the process of localization of the 'hot spot' in the lungs. A way to eliminate this effect is proposed by separate normalization of the count rates of each energy peak.

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Following the as low as reasonably achievable (ALARA) practice, there is a need to measure very low doses, of the same order of magnitude as the natural background, and the limits of detection of the dosimetry systems. The different contributions of the background signals to the total zero dose reading of thermoluminescence dosemeter (TLD) cards were analysed by using the common basic definitions of statistical indicators: the critical level (L(C)), the detection limit (L(D)) and the determination limit (L(Q)). These key statistical parameters for the system operated at NRC-Negev were quantified, based on the history of readings of the calibration cards in use. The electronic noise seems to play a minor role, but the reading of the Teflon coating (without the presence of a TLD crystal) gave a significant contribution.

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The virtual point detector assumption was validated for planar and semi-planar detectors. The distance of the virtual point detector from the detector face is larger than the thickness of the detector, which means that the virtual detector position is outside the volume of the detector.

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Considerable errors in the activity determination in lungs can be induced for the case of a "hot spot". Modern lung counter systems use several HPGe detectors, and the count rate ratios of the detectors can be used to locate the "hot spot" and apply correction algorithms. Some criteria for location determination of a point source in the lungs were investigated, and it is shown that an average error of up to about 10% can be achieved.

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The recoil fragments from an alpha-emitting source may cause significant detector contamination. The simplest method to avoid the problem is to use a source coating. The influence of Mylar film coatings of different thickness on the resolution, efficiency and minimum detection level of an alpha spectrometry system was investigated. No significant deterioration of the system characteristics was observed when using thin films of about 0.25 mg/cm2.

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Radiation workers undergo routine monitoring for the evaluation of external and internal radiation exposures. The monitoring of internal exposures involves gamma spectrometry of the whole body (whole body counting) and measurements of excreta samples. Medical procedures involving internal administration of radioactive radionuclides are widely and commonly used. Medical radionuclides are typically short-lived, but high activities are generally administered, whereas occupational radionuclides are mostly long-lived and, if present, are found generally in relatively smaller quantities. The aim of the present work was to study the interference of some common medical radionuclides (201Tl, 9mTc, 57Co, and 131I) with the detection of internal occupational exposures to natural uranium and to 137Cs. Workers having undergone a medical procedure with one of the radionuclides mentioned above were asked to give frequent urine samples and to undergo whole body and thyroid counting with phoswich detectors operated at the Nuclear Research Center Negev. Urine and whole body counting monitoring were continued as long as radioactivity was detectable by gamma spectrometry. The results indicate that the activity of medical radionuclides may interfere with interpretation of occupational intakes for months after administration.

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The hypothesis that adolescents have an exceptionally strong need for REM sleep was tested by measuring their resistance to REM deprivation. Ten adolescents (aged 16 to 17 yr.) were compared with 12 young adults (aged 25 to 27 yr.) in a standard REM deprivation procedure. The adolescents had to be awakened significantly more times than the young adults; this is consistent with the hypothesis that adolescents have a greater need for REM sleep.

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