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[en] The response of ex-core detectors depends on the core power level, power distribution, and reactor configuration. To predict the response of ex-core detectors to the core power level and power distribution, it is thus necessary to know the spatial weighting functions for ex-core detectors which are functions of reactor configurations but are independent of the core power distribution. In this study, two-dimensional discrete ordinates adjoint transport method is introduced and the calculational models are presented to calculate the spatial weighting functions. Two kinds of spatial weighting functions, i. e., assembly wise weighting functions and core height wise weighting functions, are calculated for individual ex-core detectors which are mounted at different core elevations. The relationship of spatial weighting functions to the major reactor operating parameters is also analyzed and it is found that the soluble boron in the reactor coolant does not affect the spatial weighting functions but the reactor core power level affects the relative core height wise weighting functions. It is concluded that the discrete ordinates adjoint transport method is suitable for calculating the spatial weighting functions for ex-core detectors, because the method provides more detailed information and reduces the computing times. The calculated spatial weighting functions for ex-core detectors at a given reactor configurations can be used to predict the response of each ex-core detector to the core power distribution throughout the fuel cycle, but only at a specified power level. The comparison of calculated spatial weighting functions with the measured plant data on ex-core detector response rates is needed for the validation of the calculation method
[en] The present paper is aiming at describing the injectant behaviors from two rows of film cooling holes with opposite orientation angles. Boundary layer temperature distributions were measured in the streamwise normal plane to illustrate the interaction between the injectant from the holes in the upstream row and from the downstream row. Detailed adiabatic film cooling effectiveness distributions were also measured using thermochromic liquid crystal to investigate how well the injectant covers the film cooled surface. Four film cooling hole arrangements were considered including inline and staggered ones. And three blowing ratios of 0.5, 1.0 and 2.0 were studied. At the blowing ratio of 0.5, the injectant is centered near film cooled surface irrespective of hole configurations to show high film cooling effectiveness. When the blowing ratio exceeds unity the injectant tends to lift off from the wall due to the increase of the wall normal momentum. With the inline configuration, not like other configurations, the injectant is still well attached to the wall with the help of the downwash flow formed at the hole exit
[en] A review has been made for the previous studies on safety of a geologic repository for High-Level radioactive Wastes (HLW) related to autocatalytic criticality phenomena with positive reactivity feedback. Neutronic studies on geometric and materials configuration consisting of rock, water and thermally fissile materials and the radionuclide migration and accumulation studies were performed previously for the Yucca Mountain Repository and a hypothetical water-saturated repository for vitrified HLW. In either case, it was concluded that it would be highly unlikely for an autocatalytic criticality event to happen at a geologic repository. Remaining scenarios can be avoided by careful selection of a repository site, engineered-barrier design and conditioning of solidified HLW. Thus, criticality safety should be properly addressed in regulations and site selection criteria. The models developed for radiological safety assessment to obtain conservatively overestimated exposure dose rates to the public may not be used directly for the criticality safety assessment, where accumulated fissile materials mass needs to be conservatively overestimated. The models for criticality safety also require more careful treatment of geometry and heterogeneity in transport paths because a minimum critical mass is sensitive to geometry of fissile materials accumulation
[en] In this study, we proposed a bio-hydrogen system using waste heat and by-product gas from steelworks. Dynamic simulations were carried out for the system using intermittent skirt cooling water as the heat source. The temperature changes in the components over time were compared with the measured data. The simulation results showed that the trend of temperature variation with time of each component was well predicted. Experiments and simulations agreed well not only with the design conditions but also with two off-design conditions. After confirming the reliability, we designed a 10-ton reactor for installation in the actual site using simulations. Simulation results showed that the minimum volume of the system was determined such that the effectiveness of the heat storage tank and minimum variation in the temperature of the reactor were ensured.
[en] In this research, numerical analysis was carried out on novel and existing fins, adjusted in terms of factors such as length, spacing, and angle, of a high-temperature heat exchanger for a 1 kW class Stirling engine, designed as a prime mover for a domestic cogeneration system. The performance improvement as a result of shape optimization was confirmed with numerical analysis by including the air preheater, which was not considered during optimization. However, a negative heat flux was observed in the cylinder head portion. This phenomenon was clarified by analyzing the exhaust gas and wall surface temperature of the combustion chamber. Furthermore, assuming an ideal cycle, the effects of heat transfer enhancement on the thermodynamic cycle and system performance were predicted.
[en] In this study, we compare the mass release rates of radionuclides (1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. The total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fission product radionuclides, 129I, 79Se, and 36Cl, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would be sited. The footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal
[en] Field tests of hybrid desiccant cooling systems were conducted from July to August 2011. Data were monitored and transferred in real time over the Internet. The monitored variables were analyzed to determine the performance characteristics under outdoor conditions. A series of system simulations has been conducted for outdoor conditions of the field tests. The results agree well with the experimental data in general. The system performance has been shown to deteriorate for wetter conditions, as predicted by the simulation
[en] Performance assessment of each barrier consisting of geologic disposal system for high-level radioactive wastes is carried out quantitatively, and key radionuclides and parameters are pointed out. Chemical compositions and solubilities of radionuclides under repository conditions are determined by PHREEQE code staring from compositions of granitic groundwater observed in Japan. Glass dissolution analysis based on mass transfer theory and precipitation analysis have been done in order to determine the inner boundary condition for radionuclide diffusion through a bentonite-filled buffer region, where multi-member decay chain and isotopic sharing of solubility at the inner boundary are considered. Natural barrier is treated as homogeneous porous rock, or porous rock with infinite planar fractures. Performance of each barrier is evaluated in terms of non-dimensionalized hazard defined as the ratio of annual radioactivity release from each barrier to the annual limit on intake. At the outer edge of the engineered barriers, 239Pu is the key unclide to the performance, whereas at the exit of the natural barrier, weakly-sorbing fission product nuclides such as 135Cs, 129I and 99Tc dominate the hazard. (author) 50 refs
[en] Radiation heating rates to the instruments located in the guide tubes of fuel assembly are required for the cooling analysis in the field of reactor core thermohydraulics. Previous evaluation of heating rates for the ABB-CE type power plants has been used for the design data for KSNP. But there are some differences in the structures and component materials of the instruments between them. So, it is necessary to re-evaluate the heating rates for the instruments using up-to-date cross-section library and transport code to see whether the previous evaluations are suitable for KSNP or not. The evaluations of heating rates in each component of the instruments have been performed by MCNP code and the results have been compared to those of previous works
[en] The dry new fuel storage rack shall maintain the subcritical condition (i.e., k-eff < 0.95) when fully flooded with water and the k-eff will not exceed 0.98 even assuming that the optimum moderation causes the highest reactivity. Thus, the design parameters of the new fuel rack are determined optimally by considering both the full density water flooding condition and the optimum moderation condition. The behavior of the keff as the variation of design parameter of the new fuel rack was investigated as the function of the moderating water density