NaI Scintillation Survey Meter Research Articles

Effectiveness of shielding materials against 177Lu gamma rays and the corresponding distance relationship.

The purpose of this study is to determine the dose reduction of different shielding materials at various distances from a 177Lu photon radiation source. Two protective aprons with lead equivalent thicknesses of 0.25mm and 0.35mm and tungsten-containing rubber (TCR) were used as shielding materials. A vial containing 177Lu was sealed in a lead container so that a narrow beam went out through a 3mm-diameter hole. The dose rate was measured at distances of 0, 10, 50, 100, and 200cm from the source using a NaI scintillation survey meter to obtain the rate of dose reduction. TCR was tested with thicknesses ranging from 0.3 to 1.0mm at 0.1mm intervals and from 1.0 to 4.0mm at 0.5mm intervals. At distances of 0, 10, 50, 100, and 200cm, the dose reduction for the lead equivalent thickness of 0.25mm were 32.7%, 54.5%, 93.1%, 97.9%, and 99.6%, respectively; and for the lead equivalent thickness of 0.35mm were 53.4%, 70.6%, 95.6%, 98.9%, and 99.6%, respectively. Without any shielding, the dose rate decreased by 34.4% at 10cm and by 88.8% at 50cm from the radiation source. The dose reduction for the TCR thickness of 3.5mm was 89.8% at 0cm and 93.3% at 10cm. The TCR thickness of 0.4mm provided a dose reduction comparable to or greater than that of the 0.25mm lead equivalent, whereas the TCR thickness of 1.0mm or greater provided a dose reduction comparable to that of the 0.35mm lead equivalent. Achieving a reduction of 95% or more requires the 0.25mm lead equivalent for a distance of 100cm, the 0.35mm lead equivalent for 50cm, the TCR thickness of 0.3mm for 100cm, or the TCR thickness of 0.9mm for 50cm. Without wearing a protective apron, a reduction of approximately 95% is observed at distances greater than 100cm. These findings would be useful for medical staff engaging in related activities.

In-situ determination of residual specific activity in activated concrete walls of a PET-cyclotron room

In the decommissioning work for concrete walls of PET-cyclotron rooms, an in-situ measurement is expected to be useful for obtaining a contour map of the specific activity on the walls without destroying the structure. In this study, specific activities of γ-ray-emitting radionuclides in concrete walls were determined by using an in-situ measurement method employing a portable Ge semiconductor detector, and compared with the specific activity obtained using the sampling measurement method, at the Medical and Pharmacological Research Center Foundation in Hakui, Ishikawa, Japan. Accordingly, the specific activity could be determined by the in-situ determination method. Since there is a clear correlation between the total specific activity of γ-ray-emitting radionuclides and contact dose rate, the specific activity can be determined approximately by contact dose-rate measurement using a NaI scintillation survey meter. The specific activity of each γ-ray-emitting radionuclide can also be estimated from the contact dose rate using a NaI scintillation survey meter. The in-situ measurement method is a powerful tool for the decommissioning of the PET cyclotron room.

Decontamination of the Activation Product Based on a Legal Revision of the Cyclotron Vault Room on the Non-self-shield Compact Medical Cyclotron.

The non-self-shield compact medical cyclotron and the cyclotron vault room were in operation for 27 years. They have now been decommissioned. We efficiently implemented a technique to identify an activation product in the cyclotron vault room. Firstly, the distribution of radioactive concentrations in the concrete of the cyclotron vault room was estimated by calculation from the record of the cyclotron operation. Secondly, the comparison of calculated results with an actual measurement was performed using a NaI scintillation survey meter and a high-purity germanium detector. The calculated values were overestimated as compared to the values measured using the NaI scintillation survey meter and the high-purity germanium detector. However, it could limit the decontamination area. By simulating the activation range, we were able to minimize the concrete core sampling. Finally, the appropriate range of radioactivated area in the cyclotron vault room was decontaminated based on the results of the calculation. After decontamination, the radioactive concentration was below the detection limit value in all areas inside the cyclotron vault room. By these procedures, the decommissioning process of the cyclotron vault room was more efficiently performed.

Evaluation of personal dose equivalent using optically stimulated luminescent dosemeters in Marumori after the Fukushima nuclear accident

The personal dose equivalent for 36 Marumori residents living in wooden houses was evaluated using optically stimulated luminescent dosemeters after the Fukushima nuclear accident. The cumulative doses to residents over 200 d from 1 September 2011 ranged from 536 to 1490 µSv. In each resident's house, the indoor and outdoor ambient dose equivalents were measured with an NaI scintillation survey meter. A linear relationship with a slope of 0.5 was found between them; however, the ratio of the indoor to the outdoor ambient dose equivalent varied from 0.36 to 1.0. A better linear relationship was observed between the personal dose equivalent and the indoor ambient dose equivalent compared with that between the personal dose equivalent and the outdoor ambient dose equivalent. These results indicate that the individual exposure dose depends on the indoor ambient dose equivalent rather than the outdoor ambient dose equivalent. These results should be very helpful in obtaining the countermeasures to reduce residents' exposure dose.

Measuring the external exposure dose in the contaminated area near the Chernobyl nuclear power station using the thermoluminescence of quartz in bricks.

We collected bricks from buildings in the heavily contaminated evacuated area of Belarus in a 30-km zone around the Chernobyl nuclear power station and the Gomel-Bryansk area of 150-250 km from Chernobyl and estimated the cumulative radiation dose caused by the reactor accident by measuring the thermoluminescence (TL) of the bricks. The annual dose at each location was measured using glass dosimeters and thermoluminescence dosimeters (TLD). The dose rate was measured using an energy-compensated NaI scintillation survey meter. The soil contamination near the location of each brick was measured using a germanium semiconductor detector. The main purpose of the project was to extrapolate the relation between the cumulative external dose and the present dose rate or contamination level to the lower contaminated areas. The results of the glass dosimeter, TLD, and survey meter determinations were almost identical. For a determination of the annual dose higher than 10 mGy y(-1), the cumulative dose by TL (TL dose) was roughly proportional to the annual dose and about 1.5 times larger than the cumulative dose calculated from the annual dose and 137Cs half life. The difference is expected due to the contribution of short-lived nuclides immediately after the accident or localized heavy contamination of the ground surface with 137Cs that migrated afterwards. For annual dose smaller than 10 mGy y(-1), the proportionality was not observed and most of the locations facing indoors showed TL doses very much larger than that expected from the proportionality. The cumulative dose outdoors by TL was also roughly proportional to the regional 137Cs contamination level and the proportional constant is about 10(-1) mGy per GBq km(-2), and is about 250 times larger than the present annual internal dose derived from published results. The correlation between the present dose rate where the brick was sampled and the average 137Cs concentration in the ground soil near the point is not clear, possibly because of the large spatial fluctuation in the 137Cs concentration in the soil.

Development of an Experimental System for Evaluation of the Effect of Stress on Intestinal Motility Using a Radionuclide, 51Cr, in the Rat.

We attempted to develop an experimental system for evaluation of the influence of stress on small intestinal motility using a radionuclide, 51Cr, and an acute restraint stress model. Each rat was immobilized in an adjustable restraint device for 1 to 5 h. The rat was given 51Cr immediately after the end of stress loading through a catheter inserted into the duodenum, and sacrificed 20 min after administration. The small intestine was removed and the 51Cr radioactivity was continuously monitored with an NaI-scintillation survey meter. The small intestinal motility was estimated by the distance of the radioactive peak from the top of the duodenum. The small intestinal motility was significantly inhibited by 3 h-restraint stress. This experimental system is easy, rapid, and simple for the evaluation of the effect of stress on small intestinal motility.