Performance of a concrete vault as an engineered barrier for low-level radioactive wastes was studied. For this purpose, 137Cs solution was stored in the concrete vault for approximately 600 days and the activity profile of 137Cs was measured. The profile was explained by diffusion mechanism and the apparent diffusion coefficients of 137Cs in concrete were calculated. The results suggest that there are two processes in the diffusion in concrete. The diffusion from the surface and that from the solution. The calculated apparent diffusion coefficients for each process are 1.7×10-14m2/s and 1.7×10-13m2/s respectively. Furthermore, no big difference was seen in the values for the carbonated concrete, therefore, long term barrier performance of the concrete vault is expected.
The toxicity of Ca-DTPA (calcium diethylenetriaminepentaacetic acid), Ca-EDTA (calcium ethylenediaminetetraacetic acid) and CBMIDA [Catechol-3, 6-bis (methyleiminodiacetic acid)] after intravenous injection were examined in beagle dogs. In the test (1), the change of serum total and ionic calcium levels following intravenous injection of each chelating agent was examined. The results showed that both total and ionic calium levels did not change at all by any of these chelating agents. In the test (2), each of these chelating agents (150μmol/kg) were intravenously injected to dogs daily for 1 month. No clinical untoward signs were observed in all groups. However, the significant decrease of mean body weights in Ca-DTPA groups was seen at 4 weeks after the beginning of injection. In a serum biochemical examination, the increases of GOT and GPT values in one dog, and BUN and creatinine values in another one were seen in Ca-DTPA group. No changes in all measured items were seen in Ca-EDTA group. The increases of GPT, ALP and creatinine values in one dog and GPT value in another one were observed in CBMIDA group. The slight congestion of mucosa in small intestine, the increase of lymphocytes in the lamina propria of small intestine and in the proximal tubles of kidney, and the degeneration of liver were observed in all groups. There were almost no differences in the degree of these histological changes in each group.
A practical conversion factor to estimate the value of effective dose equivalent rate in Sv unit from absorbed dose rate in air in Gy unit was examined for natural gamma radiations. The experimental examination was carried out by two methods; one measures the effective dose equivalent rate directly by using a measuring instrument having effective dose equivalent response for isotropic gamma radiations and the other obtaines it from calculation applying the gamma flux-to-effective dose equivalent factor to actual gamma energy spectrum measured in various indoor and outdoor places. From these investigations the value of the quotient of effective dose equivalent to absorbed dose in air was found do be 0.748±0.007 Sv per Gy for natural radiation exposures in various environments. The value of the quotient 0.7, which is adopted to applied to environmental gamma radiations in the UNSCEAR 1982 and 1988 Reports, was clarified to be about 7% lower than the one obtained experimentally for natural gamma radiations.
Solid state nuclear track detectors (SSNTD) record radiation in the form of tracks. In the case of high track density, however, it is not always possible to distinguish each track separately. The track density might then be underestimated unless the loss of track number due to overlapping is compensated. An elaborated “erosion” or curve fitting process is applied usually, for the separation of the overlapping tracks, to automatic track counting systems. This paper shows a much simpler correction method which was developed by the analogy of the correction equation for the dead time of GM counters. From a set of about 10 data obtained from high track density detectors, the equation for SSNTD can be determined by a least square fitting. Once the equation is found, true track density could be derived easily without any help of complex image processing or calculation, such as the erosion or curve fitting. This method also provides the privilege of extending the upper bound of the detection limit.
According to the concept of ICRP Pub. 26, the actual procedure to estimate the effective dose equivalent at Japanese nuclear power stations was studied. The study included: (1) the development of the specific instrument designed to determine the dose equivalent at a depth of 10mm of the ICRU sphere, (2) specification tests about the energy dependence, accuracy etc. on the instrument, (3) the radiation survey at standard Japanese BWR and PWR comparing the developed instrument and the standard gamma radiation exposure (rate) meter, (4) the confirmation of gamma spectra at the power stations by a spherical 3″ NaI(T1) gamma radiation monitor. It was confirmed that the ratio of the dose equivalent rate at 10mm depth and the gamma radiation exposure rate, in mSv·h-1/mR·h-1, was nearly 1/100 under the conditions of the normal operations and the outages of BWR and PWR. Therefore, it was clear that the gamma radiation exposure (rate) meters were satisfactory to determine the dose equivalent at 10mm depth of the ICRU sphere in the nuclear power stations.
For the purpose of the calibration and the energy response test, etc. of radiation measuring instruments in a low-energy region (8-75keV), the calibration fluorescent X-ray generator has been constructed to provide the suitable fluorescent X-ray calibration field. And further, its characteristics have been studied. The fluorescent X-ray spectra of radiators of Cu, Mo, Sn, Er, W, Au and Pb were measured with a high purity Ge detector. The fluorescent X-ray exposure rate were measured with standard ionization chamber. Maximum exposure rate at 80cm from radiator is about 2.58×10-5C/kg/h (100mR/h) when Sn and Ag are used as the radiator and filter respectively. The influences of scattered X-ray from radiator on the fluorescent X-rays exposure rate was less than 10%.
Study objective: To propose measures for radiological protection of veterinary workers in Japan. Design: Assessments of X-ray exposure of workers in typical conditions in veterinary clinics. Method and main results: Dose rates of useful beam and scattered radiation, worker exposure doses at different stations, and effectiveness of protective clothing were determined using TLD and ion chambers. As precautions against radiation, the following practices are important: (1) use of suitable and properly maintained X-ray equipment, (2) proper selection of safe working stations, (3) use of protective clothing. Conclusions: (1) Regulations are necessary to restrict the use of X-rays in the veterinary field. (2) Because the use of X-rays in the veterinary field is not currently controlled by law, the above precautions are essential for minimizing exposure of veterinary staff.