• Natural Product Sciences
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• v.28 no.4
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• pp.194-200
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• 2022

Albizia julibrissin Durazz. (AJ; family Minosaceae) is widely distributed worldwide, and its stem bark has been used as a traditional herbal medicine. Acute kidney injury (AKI) is a clinical syndrome that results in sudden loss of renal function. This study aimed to investigate the effects of AJ against cisplatin-induced AKI using a human kidney proximal tubule epithelial cell line (HK-2) and cisplatin-treated mice. In vitro, cisplatin treatment increased apoptosis in HK-2 cells. However, AJ treatment decreased apoptosis of cisplatin-treated HK-2 cells. In vivo, cisplatin treatment accelerated renal injury by increasing the levels of renal injury markers, such as blood urea nitrogen, creatinine, kidney injury molecule 1, and neutrophil gelatinase-associated lipocalin, which were reversed by AJ treatment. Histopathologically, AJ treatment resulted in decreased renal damage with less tubular necrosis and brush border desquamation compared with the AKI group. Additionally, cisplatin treatment upregulated mitochondrial fission, a pathological characteristic of AKI, which was downregulated by AJ treatment. Along with increased mitochondrial fission, AJ treatment also reduced cisplatin-induced apoptosis. These results suggest that AJ may be a potential therapeutic agent for cisplatin-induced AKI.

• Nuclear Engineering and Technology
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• v.13 no.4
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• pp.207-220
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• 1981

The Np-237 fission cross-sections up to 20 MeV incident neutron energy are calculated by means of the computer code STAPRE- a statistical model code with consideration of preequilibrium decay. The higher chance fissions up to third compound nucleus are taken into account, and the main input parameters in the treatment of fission under consideration of a double-humped fission barrier are carefully adjusted, so that the current trend of experimental data can be fitted within an apparent deviation of about 10% throughout the entire energy range. Results are presented in the form of point-wise cross-section values, and also in the form of graph to demonstrate the shape agreement.

• Journal of Microbiology and Biotechnology
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• v.34 no.7
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• pp.1425-1432
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• 2024

This study analyzed the effects of Ca²⁺ metal ions among culture medium components on the Chlorella sorokiniana strain DSCG150 strain cell growth. The C. sorokiniana strain DSCG150 grew based on a multiple fission cell cycle and growth became stagnant in the absence of metal ions in the medium, particularly Ca²⁺. Flow cytometry and confocal microscopic image analysis results showed that in the absence of Ca²⁺, cell growth became stagnant as the cells accumulated into four autospores and could not transform into daughter cells. Genetic analysis showed that the absence of Ca²⁺ caused upregulation of calmodulin (calA) and cell division control protein 2 (CDC2_1) genes, and downregulation of origin of replication complex subunit 6 (ORC6) and dual specificity protein phosphatase CDC14A (CDC14A) genes. Analysis of gene expression patterns by qRT-PCR showed that the absence of Ca²⁺ did not affect cell cycle progression up to 4n autospore, but it inhibited Chlorella cell fission (liberation of autospores). The addition of Ca²⁺ to cells cultivated in the absence of Ca²⁺ resulted in an increase in n cell population, leading to the resumption of C. sorokiniana growth. These findings suggest that Ca²⁺ plays a crucial role in the fission process in Chlorella.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.1
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• pp.1-7
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• 2024

In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM^-1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3775-3784
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• 2024

In this study, a skewness estimation method (SEM) and kurtosis estimation method (KEM) are introduced to determine the number of inactive cycles in Monte Carlo eigenvalue calculations. The SEM and KEM can determine the number of inactive cycles on the basis that fully converged fission source distributions may follow normal distributions without asymmetry or outliers. Two convergence criteria values and a minimum cycle length for the SEM and KEM were determined from skewness and kurtosis analyses of the AGN-201K benchmark and 1D slab problems. The SEM and KEM were then applied to two OECD/NEA slow convergence benchmark problems to evaluate the performance and reliability of the developed methods. Results confirmed that the SEM and KEM provide appropriate and effective convergence cycles when compared to other methods and fission source density fraction trends. Also, the determined criterion value of 0.5 for both ε₁ and ε₂ was concluded to be reasonable. The SEM and KEM can be utilized as a new approach for determining the number of inactive cycles and judging whether Monte Carlo tally values are fully converged. In the near future, the methods will be applied to various practical problems to further examine their performance and reliability, and optimization will be performed for the convergence criteria and other parameters as well as for improvement of the methodology for practical usage.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.90-99
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• 1994

One of the major phenomena occurring in defective fuel rods is the oxidation of UO$_2$ fuel pellets from UO$_2$ to UO$_{2+}$x/ by the oxygen Produced from the dissociation of the steam in the Pellet-to-clad gap, which leads to the enhancement of fission gas release. In this paper, the oxidation kinetics of defective fuel rods was analyzed on the basis of operating conditions of the reactor and defective fuel rod itself. Oxidation kinetics of the fuel pellet was determined under the assumption that the gap is filled with the saturated steam of 150 atm and an enhancement factor for fission gas release was introduced to take into account the effect of fuel oxidation on fission gas release. Comparison with experimental data shows that the enhancement factor predicts well the increased fission gas release due to the oxidation of UO$_2$fuel pellets.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.499-508
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• 2011

The paper briefs a fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of fuel, up to a high burnup. Of particular concern are the models for the fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in $UO_2$ fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS's precision for predicting the in-pile behavior of MOX fuel. The COSMOS code also extends its applicability to the instrumented fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code's prediction. The database consists of the $UO_2$ irradiation test up to an ultra-high burnup, power ramp test of MOX fuel, and instrumented MOX fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and $UO_2$ fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both fuel vendors for fuel design, and license organizations for an understanding of fuel in-pile behaviors.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.715-722
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• 2009

A model to predict failure of designed-to-fail (dtf) fuel particles is discussed. The dtf fuel under study consisted of a uranium oxycarbide kernel coated with a single pyrocarbon seal coat. Coating failure was assumed to be due to fission gas recoil and knockout mechanisms and direct diffusive release of fission gas from the kernel, which acted to increase pressure and stress in the pyrocarbon layer until it ruptured. Predictions of dtf fuel failure using General Atomics' particle fuel performance code for HRB-17/18 and HFR-B1 irradiation tests were reasonably accurate; however, the model could not predict the failure for COMEDIE BD-1. This was most likely due to insufficient information on reported particle fuel failure at the beginning.