ENHANCED RADIOACTIVE CONTENT OF 'BALANCE' BRACELETS.

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.

Fast determination of 9°Sr/9°Y activity in milk by Cherenkov counting.

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.

Determination of 90Sr-90Y activity in urine samples by using Cherenkov counting.

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.

A comparative study of color quenching correction methods for Cerenkov counting.

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.

A statistical model for reconstruction of historical internal doses for uranium workers.

Uranium workers are monitored for their internal doses mainly by urine measurements. During the years before the early nineties, urine samples were analysed using a fluorimetric system, and the lower limit of detection of the urine samples was relatively high, thus most of the urine measurements were recorded as below threshold. A model was developed for the reconstruction of doses to workers at uranium facilities during these years. The model is based on the assumption that the results of urine measurements are log-normally distributed and that a normalised log-normal distribution with a constant geometric standard deviation characterises the results distribution in all the uranium workplaces for workers throughout their occupational work. Therefore, the average of the urine results can be calculated from the general normalised distribution. The intake and dose can be evaluated from the reconstructed average urine concentration by assuming a constant chronic intake regime throughout the whole period of monitoring.

Color quench correction for low level Cherenkov counting.

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.

A robotic manipulator for handling TLD badges.

A prototype system for automatic handling of Harshaw/Bicron (now ThermoFisher Scientific) thermoluminescent dosimetry (TLD) badges, which is based on a robotic arm, was designed and built. The robot performs the loading and unloading of the TLD cards in the badges and transports them between the loading/unloading station and magazine stations. For quality assurance, a sticker containing the worker's details printed in barcode format was added to the badge. Automatic on-line identification is performed for checking the correlation between the badge and the TLD card number.

Determination of the solubility and size distribution of radioactive aerosols in the uranium processing plant at NRCN.

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).

Normalisation of spot urine samples to 24-h collection for assessment of exposure to uranium.

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.

Determination of radioiodine activity in charcoal cassettes.

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.

An improved method for Organic Bound Tritium (OBT) determination in urine samples.

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.

Pre- and post-irradiation fading of 6LiF:Mg,Ti (TLD-600) exposed to thermal neutrons.

The results of (6)LiF:Mg,Ti (TLD-600) glow peaks fading after irradiation by thermal neutrons, as a function of pre- and post-irradiation times up to 3 months, are presented. The measured glow curves were analysed using the computerised glow curve deconvolution and region of integration methods, to study the decay characteristics of each individual peak and the sum of selected peak areas in the glow curve. The results confirm a high stability for the sum of peaks 4 + 5 and 6 + 7, whereas peaks 2 and 3 have a dominant contribution to the fading effect.

Use of statistical checks as maintenance tools for TLD readers.

Although the values of different parameters may remain within permissible limits during the operation of a thermoluminescent dosemeter (TLD) reader, certain effects can become apparent only when a long-term followup of these parameters is performed. In order to ensure an accurate and reliable operation of a TLD reader, the system characteristics must be monitored continuously. Long-term statistical checks of key system parameters may give a broader insight into the operational characteristics of the TLD reader and may help for proper maintenance of the system. The photomultipliers noise, the internal reference light source stability and the A to D reference voltage were found to be critical parameters, which have a major influence on the accuracy and stability of the system. A followup of these parameters for a period of about 10 y is presented, and some problems are seen to be reflected in the distributions.

On the reassessment of thermal neutron doses in TLD-100 by measuring the residual dose.

By employing second readouts and the Phototransferred thermoluminescence (PTTL) method, high doses may be reassessed on the basis of residual dose information. It was shown in the past that for TLD-100, gamma doses can be reassessed by using a simple and efficient method, which consists of expanding the heating time to 30 s. In the present study, the 'extended time' method and the PTTL residual dose evaluations are used for reassessing thermal neutron doses when using TLD-100 crystals. Reassessment characteristics are presented for relatively low thermal neutron doses, in the range between approximately 1 and 18 mSv gamma dose equivalent.

Optimisation of the readout parameters when evaluating thermal neutron doses by TL dosimetry with LiF:Mg,Ti.

The location of the glow peaks depends on the heating rate. It takes some time until the crystal reaches the heater temperature, and this time lag has a direct effect on the shift of peaks towards higher temperatures. Some information on the high-temperature peaks may be lost if the readout conditions (mainly length of time) are not properly chosen. Step heating profiles to a varying final temperature between 300 degrees C and 125 degrees C were used to study the time dependence of the collected information in the glow curves of (6)LiF:Mg,Ti crystals, and the minimal heating time for evaluation of thermal neutron doses was determined.

On the reassessment of doses in TL-dosimetry by measuring the residual dose.

Second readouts and the photo-transferred thermoluminscence (PTTL) method are sometimes used to reassess high doses. When using the common LiF:Mg,Ti, if the second readout is performed by a regular readout cycle of 13.3 s, its efficiency is low and the estimations cannot be obtained with acceptable accuracy for low doses in the 10-100 mSv range. By applying the PTTL method, the efficiency is much higher, but a high background is also present, deteriorating the quality of the reassessment. A simple and efficient method was studied, which consists of expanding the heating time to 30 s. Although the efficiency relative to a standard readout is improved by only a factor of 3, the low background enables to obtain results with the same uncertainty as the more complicated PTTL method. By applying region of integration discrimination, the errors can be further diminished.

MDA improvement technique for lung counter measurements of uranium workers.

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.

Improved localization of a 'hot spot' in the lungs for an array of four HPGe detectors--the isotopic ratio effect.

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.