Nuclear engineering and technology

Liquid alloy systems have a high degree of thermal conductivity, far superior to ordinary nonmetallic liquids and inherent high densities and electrical conductivities. This results in the use of these materials for specific heat conducting and dissipation applications for the nuclear and space sectors. Uniquely, they can be used to conduct heat and electricity between nonmetallic and metallic surfaces.

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This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)- 6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice.

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This paper develops an original framework that mathematically expresses the coupling of the weight window and source biasing techniques, allowing the authors to explore the impact of inconsistent source sampling on the variance of MC results. A numerical experiment supports this new framework and suggests that certain classes of problems may be relatively insensitive to inconsistent source sampling schemes with moderate levels of splitting and rouletting.

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Mechanical shim is an advanced technology for reactor power and axial offset control with control rod assemblies. To address the adverse accuracy impact on the ex-core power range neutron flux measurements-based axial offset control resulting from the variable positions of control rod assemblies, the lead-lag-compensated in-core self-powered vanadium detector signals are utilized.

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The focus on the issues surrounding spent nuclear fuel and lifetime extension of old nuclear power plants continues to grow nowadays. A transparent decision-making process to identify the best suitable nuclear fuel cycle (NFC) is considered to be the key task in the current situation.

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Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation.

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The adsorption of uranium (VI) by calcium alginate beads was examined by batch experiments. The effects of environmental conditions on U (VI) adsorption were studied, including contact time, pH, initial concentration of U (VI), and temperature. The alginate beads were characterized by using scanning electron microscopy, transmission electron microscopy, X-ray photoelectron spectroscopy, and Fourier transform infrared spectroscopy. Fourier transform infrared spectra indicated that hydroxyl and alkoxy groups are present at the surface of the beads.

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Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

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A dynamic nanoindentation method was applied to study an Fe-18 at.% Gd alloy as a neutron-absorbing material prepared by vacuum arc-melting and cast in a mold. The Fe-18 at.% Gd cast alloy had a microstructure with matrix phases and an Fe-rich primary dendrite of Fe9Gd. Rietveld refinement of the X-ray spectra showed that the Fe-18 at.% Gd cast alloy consisted of 35.84 at.% Fe3Gd, 6.58 at.% Fe5Gd, 16.22 at.% Fe9Gd, 1.87 at.% Fe2Gd, and 39.49 at.% b-Fe17Gd2. The average nanohardness of the primary dendrite phase and the matrix phases were 8.7 GPa and 9.3 GPa, respectively.

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Three plastic scintillators of 4.5 cm diameter and 2.5-cm length were fabricated for comparison with commercial plastic scintillators using polymerization of the styrene monomer 2.5-diphenyloxazole (PPO) and 1,4-bis benzene (POPOP). Their maximum emission wavelengths were determined at 426.06 nm, 426.06 nm, and 425.00 nm with a standard error of 0.2% using a Varian spectrophotometer (Agilent, Santa Clara, CA, USA).

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In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to UO3, then reduced to UO2 and hydroflorinated to UF4. Effect of two different process routes of ADU precipitation on the characteristics of ADU, UO3, UO2 and UF4 were studied here.

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The evaporation of I2 was observed to follow primarily first order kinetics, depending on the I2 concentration. The evaporation rate constant increased rapidly with increase in temperature. The presence of Ie slightly reduced the evaporation rate of I2 by forming relatively stable I 3 . The effect of Cle at

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The measured Nusselt numbers showed a trend similar to those of existing studies, but the absolute values showed discrepancies owing to the high Prandtl number of this system. The measured angle-dependent Nusselt numbers were analyzed for 2-D and 3-D geometries, and a multiplier was developed that enables the extrapolation of 2-D data into 3-D data.

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The noise level of the temperature measurement system was approximately 0.08 mK (peak to peak). The temperature of the core part rose by approximately 8.6 mK at 800 MU (monitor unit) for 6-MV Xray beams, and it increased as X-ray energy increased. The temperature rise showed less spread when it was normalized to the accumulated charge, as measured by an external monitoring chamber.

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The results show that the medical staff effective dose is highly dependent on the position of the medical staff. The results also show that the maximum medical staff effective dose in an emergency situation in the presence of a patient is ~25.5 mSv/s.

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The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels.

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This work presents a new technique for discriminating between alpha particles of different energy levels. In a first study, two groups of alpha particles emitted from radium-226 and americium-241 sources were successfully separated using a CR-39 microfilm of appropriate thickness. This thickness was adjusted by chemical etching before and after irradiation so that lower-energy particles were stopped within the detector, while higher-energy particles were revealed on the back side of the detector.

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In this study, we compared a PCD and conventional scintillation detector with respect to the energy windows (5%, 10%, 15%, and 20%) using a 99mTc gamma source with a Geant4 Application for Tomography Emission simulation tool. The gamma camera systems used in this work are a CZT PCD and NaI(Tl) conventional scintillation detector with a 1-mm thickness. According to the results, although the IU and DU results were improved with the energy window, the SF and CNR results deteriorated with the energy window.

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The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated.

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This article suggests a simple parameterization for the gas bubble rising velocity as a function of the volume-equivalent bubble diameter; this parameterization does not require prior knowledge of bubble shape. This is more convenient than previously suggested parameterizations because it is given as a single explicit formula.

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