To strengthen the thin window of a proportional gas flow counter which is capable to detect low energy β-ray such as tritium is described. The window film is made of alkyl-benzene whose thickness is extremely thin (0.15mg/cm2) and has large surface (30mm×150mm). In order to strengthen the window without increasing the thickness of the film, a latticed frame was installed to support the window film. Calculation for approximating of strengthening of the thin window was discussed. Mechanical strength of the window was estimated to withstand a pressure difference 66 times greater than that without the frame. It was also proved that the strength of the film by the pressure difference calculated agreed with that of experimental results with the accuracy of 10%. It is confirmed that the gas flow counter for measuring low energy β-ray developed in our laboratory, inspite of using the latticed frame, has identical counting efficiency without latticed frame.
The rates of photolysis by sunlight of radioactive methly iodide released from a nuclear facility to the atmosphere were estimated theoretically. The photolysis rates of methyl iodide were also measured experimentally using stable methyl iodide gas which was irradiated by sunlight in the reaction vessel. The experimental values were compared with the calculated ones. On a clear day in the low level of air pollution, the hourly maximum photolysis rate estimated is about 3.6%hr-1 at noon in the early of July and 1.5%hr-1 in winter. In annual variation of the daily photolysis rate on a clear day, the maximum (about 24% day-1) appears in July and the minimum (about 7% day-1) in December. The half life of methyl iodide by photolysis is estimated to be about 3 days in summer and 9 days in winter. The calculated values are in agreement with the experimental ones within a factor of 2.
A practical method for assessing ‘shallow dose equivalent’ for beta-rays, as secondary limits of protection standards, has been studied using Monte Carlo method. Dose distribution at various depths within a surface layer of human body were estimated in the cases of irradiation from 3H, 14C, 147Pm, 133Xe, 131I, 85Kr, 204Tl, 198Au, 133I, 32P, and 90Sr-90Y. We used a model, such as slab-receptor exposed to beta-rays under conditions of broad parallel beam. The point of the maximum dose equivalent within the shell from 7mg/cm2 to 1, 000mg/cm2 in the slab-receptor was located at the depth of 7mg/cm2. We suggest to make use of dose equivalent at the depth of 7mg/cm2 as secondary limits to prevent non-stochastic effects on the skin and the lens, and to define it shallow dose equivalent. Moreover the conversion factors for the shallow dose equivalent to receptor-free air dose for beta-ray of energies (Emax) below 2.245MeV were examined. We proposed that the conversion fatter for the beta-rays is taken to be 1.1.
The measurement of fast neutron flux was made by means of electrical spark counting of etched nuclear tracks on polycarbonate film (Makrofol KG) of 10μm thickness. 232Th electrodeposited disk was used as a detector target. By this method the upper limit of spark counts was about 660/cm2 of detector film. A linear relation was obtained from 5×106 to 4×1010n/cm2 of neutron fluence for 199μg/cm2 thorium target. Detection sensitivity (the ratio of spark counts per unit area to fast neutron fluence) was estimated to be 3.8×10-8 counts/n, hence one spark count equivalent to 189 mrem (1.89mSv). Process of thorium electroplating on a stainless steel disk as a target for fast neutron detection was described in detail. Thorium nitrate dissolved in 0.035M ammonium oxalate [(NH4)2C2O4] was used as electrolyte. Maximum electroplating efficiency obtained was about 70% with electric current density 22mA/cm2 and pH 6-8 at 80°C for 3hr.
Two methods have been developed for rapid determination of 89Sr and 90Sr in liquid effluent from nuclear facilities. One is the sequential Cerenkov and liquid scintillation counting. The pure β emitting nuclides, freshly separated from 90Y by 0.05M TTA-Benzene extraction, are first counted for 89Sr Cerenkov radiation, and then recounted for 89Sr and 90Sr by liquid scintillation counting. The other is the simultaneous determination of 89Sr and 90Sr with spill-over method by liquid scintillation counting. Although ingrowing 90Y may be a main cause of errors in determination of 89Sr, it can be minimized by dual radionuclides counting in Cerenkov and triple radionuclides counting in liquid scintillation. Reasonably accurate results have been obtained for samples having 89Sr/90Sr ratio ranging from 0.05 to 20, if the countings are completed within several hours after chemical separation of 90Y.