Making use of the radioisotopes of manganese, zinc and antimony, to represent radioactive corrosion products, and a radioisotope of cesium as representative fission product, the transport and deposition of these radioisotopes in a natural circulation sodium loop have been studied over a temperature range from 630°C down to 280°C. The movement of the radioisotopes during the loop running were monitored by in situ counting and radiochemical analysis of sodium samples taken from the loop. The distribution of the radioisotopes over the various cross sections along the cooling tube were determined after quick solidification of flowing sodium by chilling. The results indicate that manganese deposits on the substrate materials presumably by adsorption in the higher-temperature region and by precipitation in the lower-temperature region, zinc presumably by adsorption associated with plating in the lower-temperature region, antimony by coprecipitation with other deposits in the cold settling pocket, and cesium mainly by physical adsorption in the lower-temperature region. Significant deposition of the radioisotopes was observed in the pocket and the apparent rate constants of the deposition were obtained for manganese, zinc and cesium.
Numerical calculations have been made of both noise spectra and relative standard deviations for stochastic fluctuations in various state quantities of a system, such as for example, neutron number, coolant temperature and coolant flow-speed. The calculations are based on a theoretical model proposed earlier by the author for a non-boiling liquid-cooled and -moderated reactor, and carried out for the case of natural convection cooling at various values of reactor power up to 100kW. Some of the results are compared with experiment. It is shown that the low-frequency fluctuations, caused by coolant flow-speed fluctuations, become significant at increased power levels, and above several kW, the fluctuations in flow are visibly reflected in those of neutron number.
The total value of the separative work provided by separators in a step cascade, in which mixing cannot be avoided, is larger than that provided by a corresponding ideal cascade operated under the same conditions (rate and concentration) of product, feed and waste. This difference in the separative work is due to mixing loss. In this study, a method for calculating the mixing loss in a step cascade is proposed, in the case where the separation factor is nearly 1. Calculations using this method reveal that in each step there exists a stage at which no mixing takes place, provided that the value of the cut in the step is nearly the value of the cut minimizing the mixing loss in the step. As the value of the cut in the step deviates from the value of the cut minimizing the mixing loss in the step, the position of the mixing free stage shifts upwards of downwards from mid-position of the step.
Removal of 60Co and 65Zn from aqueous solution in a column by the chelating resin Dowex A-1 has been studied in its relation to pH, to the concentration of sodium salts added to the influent and to the flow-rate of the influent. It is shown that the decontamination factors (DF) of 60Co and 65Zn both increase consistently with pH between 3.5 and 5.0, beyond which the DF falls back. In relation to sodium salt concentration, the DF value decreases consistently with increasing concentration in the range of 0.1∼1.0N, above which, further addition of sodium salts has no further effect. The DF value is completely independent of flow-rate of the influent in range of 16∼160hr-1 space velocity. It is, further, very little affected by differences in effluent volume, in the range from 6.7 to 1.3×103 bed volumes. For the regeneration of the resin, about 99.9% of 60Co or 65Zn is removed from the resin by at least 3 bed volumes of 0.1M EDTA or 0.1M NTA used as regenerating solution.
The Snoek peak of a neutron-irradiated Fe-3.48% Ni alloy was measured to determine the resolution of interstitial impurity atoms. The internal friction was measured after isothermal annealing on specimens irradiated at 75°±10°C to a dose of ∼3.6×1019 nvt. The change of the Snoek peak height, which corresponds to the amount of nitrogen in solution, was analyzed upon assuming the applicability of the rate equation. As a result of calculation, the order of the reaction was found to be about 2.6, and the activation energy for resolution of nitrogen atoms about 2.9eV.
The ratio of the γ-activities per fission from fission products of 239Pu and 235U, and its time dependence were measured by double fission chamber technique. The γ-activity from the fission products of 239Pu fission was lower than the corresponding activity relevant to 235U fissions. The ratio varied with the cooling time allowed after irradition. This ratio was applied to power distribution measurements by γ-scanning method in multi-region cores composed of PuO2-UO2 and UO2 fuels. To obtain the relative power, the measured γ-activities from the fission products in the fuel rods were corrected for the difference between the γ-activities per fission from the fission products.
The behavior of sodium oxide aerosol in a closed chamber was studied for the safety analysis of a Na-cooled fast reactor. The experimental apparatus and techniques are first described. The aerosol was released during a short time by blowing air onto heated Na in a 1m3 chamber. The maximum mass concentration of the aerosol in the form of Na2O ranged of 0.05∼10g/m3. The particle size distribution, the aerosol mass concentration and the mass deposition rates were measured as a function of time. It was found that the mass median diameter of the aerosol was related to the maximum mass concentration. To determine the character of the behavior of sodium oxide aerosol in the chamber, the density of the aerosol material and the thickness of the boundary layer through which the particles deposit on the chamber wall were observed. The initial half-time of the aerosol mass concentration was compared with the values numerically calculated under certain assumptions.