• Structural Engineering and Mechanics
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• v.39 no.1
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• pp.125-145
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• 2011

Pushover analysis has gained significant popularity as an analytical tool for realistic determination of the inelastic behaviour of RC structures. Though significant work has been done to evaluate the demands realistically, the evaluation of capacity and realistic failure modes has taken a back seat. In order to throw light on the inelastic behaviour and capacity evaluation for the RC framed structures, a 3D Reinforced concrete frame structure was tested under monotonically increasing lateral pushover loads, in a parabolic pattern, till failure. The structure consisted of three storeys and had 2 bays along the two orthogonal directions. The structure was gradually pushed in small increments of load and the corresponding displacements were monitored continuously, leading to a pushover curve for the structure as a result of the test along with other relevant information such as strains on reinforcement bars at critical locations, failure modes etc. The major failure modes were observed as flexural failure of beams and columns, torsional failure of transverse beams and joint shear failure. The analysis of the structure was by considering all these failure modes. In order to have a comparison, the analysis was performed as three different cases. In one case, only the flexural hinges were modelled for critical locations in beams and columns; in second the torsional hinges for transverse beams were included in the analysis and in the third case, joint shear hinges were also included in the analysis. It is shown that modelling and capturing all the failure modes is practically possible and such an analysis can provide the realistic insight into the behaviour of the structure.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.884-894
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• 2023

Design and safety assessment of fuel pins for application in innovative Generation IV fast reactors calls for a dedicated nuclear fuel modelling and for the extension of the fuel performance code capabilities to the envisaged materials and irradiation conditions. In the INSPYRE Project, comprehensive and physics-based models for the thermal-mechanical properties of U-Pu mixed-oxide (MOX) fuels and for fission gas behaviour were developed and implemented in the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. As a follow-up to the assessment of the reference code versions ("pre-INSPYRE", NET 53 (2021) 3367-3378), this work presents the integral validation and benchmark of the code versions extended in INSPYRE ("post-INSPYRE") against two pins from the SUPERFACT-1 fast reactor irradiation experiment. The post-INSPYRE simulation results are compared to the available integral and local data from post-irradiation examinations, and benchmarked on the evolution during irradiation of quantities of engineering interest (e.g., fuel central temperature, fission gas release). The comparison with the pre-INSPYRE results is reported to evaluate the impact of the novel models on the predicted pin performance. The outcome represents a step forward towards the description of fuel behaviour in fast reactor irradiation conditions, and allows the identification of the main remaining gaps.

• Tribology and Lubricants
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• v.40 no.3
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• pp.97-102
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• 2024

The next-generation Molten Salt Reactor is known for its high safety because it uses nuclear fuel dissolved in high-temperature molten salt, unlike traditional solid atomic fuel methods. However, the high-temperature molten salt causes severe corrosion in internal structural materials, threatening the reactor's safety. Therefore, it is crucial to investigate the high-temperature corrosion resistance and wear performance of materials used in reactors to ensure safety. In this study, the high-temperature corrosion resistances and wear performances of corrosion samples in a NaCl-MgCl2-KCl (20-40-40 [wt%]) molten salt are investigated to evaluate the applicability of economically viable stainless steels, 316SS and 304SS. Hastelloy C276 and a new alloy containing a small amount of Nb are used as reference samples for comparative analysis. The mass loss, mass loss rate per unit volume, and surface roughness of each sample are measured to understand the corrosion mechanisms. Scanning electron microscopy and energy-dispersive spectroscopy analyses are employed to analyze the corrosion mechanisms. Wear tests on the corroded samples are also conducted to assess the extent of corrosion. Based on the experimental results, we predict the lifespans of the materials and evaluate their suitability as candidate materials for molten salt reactors. The data obtained from the experiments provide a valuable database for structural materials that can enhance the stability of molten salt reactors and recommend high-temperature corrosion-resistant materials suitable for next-generation reactors.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2282-2291
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• 2024

In this work we study the effects of different factors of dislocation loop on its obstacle strength when interacting with an edge dislocation. At first, the interaction model for dislocation and dislocation loop is established and the full and partial absorption mechanism is obtained. Then, the effect of temperature, size and burgers vector of dislocation loop are investigated. The relation between the obstacle strength and irradiation dose has been established, which bridges the irradiation source and microscale properties. Except that, the obstacle strength of C, Cr, Ni, Mn, Mo and P decorated dislocation loop is studied. Results show that the obstacle strength for dislocation loop decorated by alloy element decreases in the sequence of Cr, Ni, Mn, C, P and Mo, which could be used to help parameterize and validate crystal plasticity finite element model and therein integrated constitutive laws to enable accounting for irradiation-induced chemical segregation effects.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1584-1602
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• 2024

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.216-221
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• 2016

Background: Very fine radiation-induced aerosol particles are produced in intense radiation fields, such as high-intensity accelerator rooms and containment vessels such as those in the Fukushima Daiichi nuclear power plant (FDNPP). Size measurement of the aerosol particles is very important for understanding the behavior of radioactive aerosols released in the FDNPP accident and radiation safety in high-energy accelerators. Materials and Methods: A combined technique using wire screens and imaging plates was developed for size measurement of fine radioactive aerosol particles smaller than 100 nm in diameter. This technique was applied to the radiation field of a proton accelerator room, in which radioactive atoms produced in air during machine operation are incorporated into radiation-induced aerosol particles. The size of $^{11}C$-bearing aerosol particles was analyzed using the wire screen technique in distinction from other positron emitters in combination with a radioactive decay analysis. Results and Discussion: The size distribution for $^{11}C$-bearing aerosol particles was found to be ca. $70{\mu}m$ in geometric mean diameter. The size was similar to that for $^7Be$-bearing particles obtained by a Ge detector measurement, and was slightly larger than the number-based size distribution measured with a scanning mobility particle sizer. Conclusion: The particle size measuring method using wire screens and imaging plates was successfully applied to the fine aerosol particles produced in an intense radiation field of a proton accelerator. This technique is applicable to size measurement of radioactive aerosol particles produced in the intense radiation fields of radiation facilities.

• Transactions of the Korean hydrogen and new energy society
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• v.26 no.5
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• pp.445-452
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• 2015

A tubular packed bed reactor for the steam-$CO_2$ combined reforming of natural gas to produce the synthesis gas of a target $H_2/CO$ ratio 2.0 was simulated. The effects of the reactor dimension, the feed gas composition, and the gas feeding temperature upon the possibility of coke formation across the catalyst bed were investigated. For this purpose, 2-dimensional heterogeneous reactor model was used to determine the local gas concentrations and temperatures over the catalyst bed. The thermodynamic potential distribution of coke formation was determined by comparing the extent of reaction with the equilibrium constant given by the reaction, $CH_4+2CO{\Leftrightarrow}3C+2H_2O$. The simulation showed that catalysts packed in the central region nearer the entrance of the reactor were more prone to coking because of the regional characteristics of lower temperature, lower concentration of $H_2O$, and higher concentration of CO. With the higher feeding temperature, the feed gas composition of the increased $H_2O$ and correspondingly decreased $CO_2$, or the decrease in the reactor diameter, the volume fraction of the catalyst bed subsequent to coking could be diminished. Throughout the simulation, reactor dimension and reaction condition for coking-free operation were suggested.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.62-70
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• 2019

The APR1400 reactor may be operated for a long time under high temperature and pressure conditions, causing damage to the stud holes and causing stud bolts and holes to stick. The present practice is to manually remove the anti-sticking agent and foreign matter remaining in the APR1400 reactor stud hole and to visually check the surface condition of the thread to check the damage status of the threads. In the case of the APR1400 reactor stud holes, manually cleaning the threads increases the risk of radiation exposure and operator's fatigue. To avoid this, the autonomous mobile robot is used to automatically clean the reactor stud holes. The purpose of this study is to optimize the cleaning performance of the mobile robot by looking at the behavior of the surface roughness of the stud surface cleaned by the brush attached to the mobile robot due to changes in brush material, thickness of wire, and rotation speed. A microscopic approach to the surface roughness of the flank is needed to investigate the effects of the newly proposed brush of the autonomous mobile robot on the thread holes. According to this experiment, it is reasonable to use STS brush rather than Carbon one. Optimal operating conditions are derived and the safety of APR1400 reactor stud holes maintenance can be improved.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.