• Journal of Radiation Protection and Research
• /
• v.7 no.1
• /
• pp.56-60
• /
• 1982

A calculation of the effective multiplication factor has been made for GODIVA and JEZEBEL critical assemblies by using a computer code, ANISN, with having the Watt's, Cranberg's and Maxwellian formulae for the prompt fission neutron spectrum as a fission source. Then the calculated values have been compared with experimental data obtained by others. The Maxwellian formula seems to be the best one for representing the prompt fission neutron spectrum since the effective multiplication factor based on it shows a better agreement with the experimental value compared to the rest formulae.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2878-2887
• /
• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

• Nuclear Engineering and Technology
• /
• v.50 no.7
• /
• pp.1060-1067
• /
• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• Biomolecules & Therapeutics
• /
• v.30 no.5
• /
• pp.409-417
• /
• 2022

Parkinson's disease (PD) is the second most common neurodegenerative disease worldwide, and accumulating evidence indicates that mitochondrial dysfunction is associated with progressive deterioration in PD patients. Previous studies have shown that sinapic acid has a neuroprotective effect, but its mechanisms of action remain unclear. The neuroprotective effect of sinapic acid was assayed in a PD mouse model generated by the neurotoxin 1-methyl-4-phenyl-1,2,3,6-tetrahydropyridine (MPTP) as well as in SH-SY5Y cells. Target protein expression was detected by western blotting. Sinapic acid treatment attenuated the behavioral defects and loss of dopaminergic neurons in the PD models. Sinapic acid also improved mitochondrial function in the PD models. MPTP treatment increased the abundance of mitochondrial fission proteins such as dynamin-related protein 1 (Drp1) and phospho-Drp1 Ser616. In addition, MPTP decreased the expression of the REV-ERB α protein. These changes were attenuated by sinapic acid treatment. We used the pharmacological REV-ERB α inhibitor SR8278 to confirmation of protective effect of sinapic acid. Treatment of SR8278 with sinapic acid reversed the protein expression of phospho-Drp1 Ser616 and REV-ERB α on MPTP-treated mice. Our findings demonstrated that sinapic acid protects against MPTP-induced PD and these effects might be related to the inhibiting abnormal mitochondrial fission through REV-ERB α.

• Natural Product Sciences
• /
• v.28 no.4
• /
• pp.194-200
• /
• 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
• /
• v.13 no.4
• /
• pp.207-220
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.729-738
• /
• 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
• /
• v.26 no.1
• /
• pp.90-99
• /
• 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
• /
• v.43 no.6
• /
• pp.499-508
• /
• 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
• /
• v.54 no.7
• /
• pp.2395-2407
• /
• 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.