Distributions of the TRU nuclide composition in spent fuels are important for radiation-shielding analyses in their storage and transport. In this research, detailed three-dimensional distributions of 242Cm and 244Cm compositions as the main neutron sources in spent fuel assemblies (BWR 8×8 and 9×9 types) were calculated with the three-dimensional neutron transport and fuel burnup code system. The distributions of 242Cm and 244Cm compositions become heterogeneous as fuel burnup increases. Horizontally, the distributions of both compositions in the fuel assemblies always show peaks at the corner rods. Vertically, in the corner rod, the distributions show peaks on the bottom nodes at the initial burnup, and the peak node gradually moves upward as fuel burnup increases until the fuel assemblies are unloaded. For example, in a 9×9 type fuel assembly at the final burnup (55 MWd/kgU), the horizontal peak ratio of the 244Cm composition was 1.78 at the corner rod and its vertical peak ratio was 1.67 on the upper node. 242Cm and 244Cm compositions mainly contribute to neutron emission rates in spent fuel. Accordingly, it is rational to consider those heterogeneous distributions for the neutron-shielding evaluation of spent fuels of higher burnup.
The enhancement of safety culture is an issue for both plant operators and regulators working in fields related to the safety management of nuclear power plants. Plant operators have been collecting safety culture data through a broad range of tools and methods such as observations, interviews and other surveys. However, many issues remain regarding the effectiveness of safety culture activity. A new regulatory authority was established in 2012 after the Fukushima Dai-ichi accident to enforce nuclear safety regulations but its activity is still weak in terms of monitoring the performance of plant operators’ safety culture. In order to promote and strengthen safety culture, plant operators need to collect detailed event information, even when events do not directly affect plant safety, and accumulate information related to human, organizational and technical factors through dialogue among the parties responsible for coping with events. Both operators and regulators should all be working in the same direction to assess information that will help their periodic reviews, and the involvement of local government is a key to enhancing their safety culture.
Among twelve FCI cases examined in the OECD/NEA/CSNI/SERENA2 test series using two facilities, six steam explosion cases, five from TROI and one from KROTOS, were analyzed by JASMINE V.3. The major model parameters were categorized into the “focused zone”, i,e., the core part of interest, and the “peripheral zone”, corresponding to the initial and boundary conditions given intentionally for each test case. For the former, base values established through past validation studies of JASMINE V.3 were applied. The code was modified to implement the measured distribution of the entrained droplet size acquired in TROI–VISU. For the latter, melt release histories were given as a combination of time tables of jet diameter and release velocity that were estimated from image data and transit timing data of the melt leading edge. The base values were shown to predict impulse responses of SERENA2 systematically with a reasonable error band. A statistical analysis based on the LHS method was performed. Uncertainty ranges were given on the basis of measurement errors and past validation studies in the development of JASMINE. Underlying mechanisms causing apparent differences in the mechanical energy conversion ratio between two facilities were studied from the viewpoint of breakup length and trigger timing.
In order to estimate the fuel behavior during a severe accident in a fast reactor, it is necessary to make pure sodium uranates and to measure their properties. In the present study, sodium uranates (NaxUyOz) were synthesized at various reaction temperatures for various durations using UO2 and Na2O or Na2CO3. We succeeded in synthesizing nearly pure sodium uranates such as Na2U2O7, α–Na2UO4, NaUO3, and Na4UO4. Information for making α–Na2UO4 and NaUO3 with higher purity was obtained. The phase transition temperature from the α–phase to the β–phase of a sample of Na2U2O7 was clarified to be 658.2 K by thermal expansion measurement.
Water contamination with radionuclides has been found to occur during the processing of fuel debris by the submersion method. In this study, we have investigated the collective removal of radionuclides by the solvent extraction method with the ultimate goal of developing a process for treating such contaminated water. Co(II), Ni(II), Sr(II), Zr(IV), Mo(VI), Cs(I), and Nd (III), which are assumed to be present in the fuel debris, are chosen as the target species for the solvent extraction process. Octyl(phenyl)–N,N–diisobutylcarbamoylmethylphosphine oxide (CMPO)–tributyl phosphate (TBP)–[C2mim][Tf2N], which serves as the extractant, exhibits a strong ability to extract Co(II), Ni(II), Sr(II), and Nd (III) at a NaCl concentration of 2.0 mol/dm3, whereas it exhibits high extraction ability towards Co(II), Ni(II), Sr(II), Zr(IV), Cs(I), and Nd(III) at a H3BO4 concentration of 0.1 mol/dm3. This extractant also quickly extracts the radionuclides, allowing extraction equilibrium to be reached within 3 min in the cases of Co(II), Ni(II), Sr(II), Zr(IV), Cs(I), and Nd(III). The results of the present study suggest that the radionuclides present in contaminated water can be extracted with high efficiency using CMPO–TBP–[C2mim][Tf2N] as the extractant.
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