Uranium glass is yellowish green glass which is produced by adding a small amount of uranium as a coloring agent into ordinary glasses. The glass has a characteristic of emitting strong fluorescent light of around 550nm under ultraviolet light. Most of uranium glass was fabricated before World War 2 but still small amount of uranium glass is being fabricated nowadays. In this study, various kinds of uranium glass were subjected to the measurement of the γ-ray by Ge semiconductor detector. It was found that in older glasses (fabricated from the late 19th century to the early 20th century), γ-ray peaks from 214Pb and 214Bi were clearly identified which are in equilibrium with 226Ra that was apparently not completely separated from uranium due to the level of purification technology at that period. From the comparison of gamma peaks of 235U and 234Pa, it was found that many uranium glass samples fabricated after World War 2 use depleted uranium. The content of 40K in uranium glass was found to vary with where specimens were produced.
A theoretical model for Alpha radioactivity Measurement using ionized Air Transportation (AMAT) is developed to predict the relation between alpha radioactivity and measured ion current. AMAT is very efficient for measuring alpha contaminated wastes with large and complex surfaces. However, since the theoretical relation between alpha radioactivity and observed ion current is unclear because of the complicated behavior of ionized air molecules, one-dimensional model covering from ionization of air molecules to ion detection is developed based on atmospheric electrodynamics. The model prediction of ion current was quantitatively in agreement with the experimentally observed ion current in the simple AMAT system, within 40% errors of absolute ion current values. Also, the model can qualitatively predict the dependency of measured ion current on the air velocity.
A detailed gas-liquid two-phase flow analysis code based on an advanced interface-tracking method has been developed. It is expected that the developed code would be able to simulate two-phase cross flow behavior within tight-lattice fuel bundles without relying on any empirical correlations. In order to verify the applicability of the code to simulate twophase cross flow behavior in such situations, numerical analyses of 2-channel model tests were conducted to compare the air slug deformation and separation behavior caused by cross flow through a narrow interconnection between channels. Although the code underestimated the ascending velocity of the slug, the calculated slug deformation and separation behavior were shown to be quite similar to those observed by a high-speed video camera. Moreover the minimum differential pressure between the subchannels through the interconnection, causing channel-to-channel air transfer to occur could be predicted to within 20Pa. However, further studies of modeling and implementation related to the interface-channel wall interaction, such as a contact angle of a gas-liquid interface at the channel wall, are required for prediction improvements. Nevertheless, the qualitative capability of the developed code to simulate two-phase cross flow phenomena was demonstrated.
The reduced-moderation water reactor (RMWR) core adopts a hexagonal tight-lattice arrangement with about 1mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermalhydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing an advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the verification works of the two-phase flow simulation code TPFIT. The numerical results applied to liquid film falling down on an inclined flat plate are shown and compared with existing experimental results. In the results of numerical simulation, the development of the wave structures could be reproduced by the numerical simulation. The predicted average local film thicknesses almost agreed with Nusselt's mean film thickness. Moreover, the probability density function of local film thickness agreed well with the experimental result. In addition, the trends related to the film velocities were consistent with the theoretical results. From these results, it was confirmed that the TPFIT code is applicable to predict the liquid film flow behavior.
In this study, the corrosion behavior of stainless steel was investigated in consideration of characteristics of reprocessing solution for spent nuclear fuel. The effects of uranium, plutonium, fission products, corrosion products etc. on nitric acid corrosion of austenitic stainless steels were evaluated to understand the corrosion factors in the reprocessing solution and were electrochemically discussed. As results, the oxidizing ions such as Pu(VI), Ru(III) and Cr(VI) accelerated the corrosion of stainless steel in hot nitric acid solutions, however uranium effect was not significant. The effectivitity of the oxidizing ions on corrosion of stainless steel is related to the standard electrode potential (E0). Roughly, the oxidant of equilibrium reaction, which have E0 more than 1.0V vs NHE, are possible to accelerate the corrosion of stainless steel in nitric acid solution. The effectivity of oxidants, which have lower E0 in the range, on stainless steel corrosion was quite significant because they are stabler in hot nitric acid solutions.
Nuclear Power Engineering Corporation (NUPEC) has performed conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program BASALA following MISTRAL. They were devoted to measuring the core physics parameters of such advanced cores. The MISTRAL program consists of one reference UO2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. As for MISTRAL, the analysis results have already been reported on April 2003. The BASALA program consists of two high moderation full MOX BWR mock-up cores for operating and cold stand-by conditions. NUPEC has analyzed the experimental results of BASALA with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. The analysis results of MISTRAL and BASALA indicate that these applied analysis methods have the same accuracy for the UO2 and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.
Recently, it has been reported that the conventional calculation method for fast neutron systems loses its validity when it is applied to fast systems that are significantly sensitive to iron cross section. In the present paper, a newly developed cell calculation code, SLAROM-UF, has been applied to calculations for such systems. SLAROM-UF utilizes the ultra-fine energy group library below 50keV and the 900-group library to estimate the self-shielding effect caused by resonances of heavy nuclides and wide resonances of structural materials, respectively. When SLAROM-UF with 900-group library was applied to cell calculation and core calculation was performed in a properly adopted 220-group structure, discrepancy of multiplication factor from the continuous energy Monte-Carlo calculation was reduced from 2.0 to 0.4%Δk. Large dependency on energy group used for core calculation is observed in JOYO MK-III, It is caused by "the fuelreflector interface effect" which is recently discussed as a problem for calculation of fast neutron systems.
De-nitration process of Pu-U nitrate solution by the microwave heating was developed for the aim at non-proliferation of nuclear materials. After measuring the accurate temperature history in the heating process by a newly developed thermometer, the reaction mechanism was investigated at four points by using TG-DTA (Thermo gravity-differential thermal analysis) and X-ray diffraction analysis. From these results, the most likely processes on the dehydration, denitration and formation of oxide for de-nitration process of UO2(NO3)2 were suggested.