• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.939-946
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• 2023

Comprehensive Research Facility for Fusion Technology (CRAFT) is an integration of different demonstrating or testing facilities, which aim to develop the critical technology or composition system towards the fusion reactor. Due to the importance and challenge of the negative ion based neutral beam injection (NNBI), a NNBI test facility is included in the framework of CRAFT. The initial object of CRAFT NNBI test facility is to obtain a H⁰ beam power of 2 MW at the energy of 200-400 keV for the pulse duration of 100 s. Inside the negative ion accelerator of NNBI system, the interactions of the negative ions with the background gas and electrodes can generate abundant stray electrons. The stray electrons can be further accelerated and dumped on the electrodes or eject from the accelerator. The stray electrons, including the ejecting electrons, cause the unwanted particle and heat flux onto the electrodes and the inner components of beamline (especially the temperature sensitive cryopump). The suppression of the stray electrons from the CRAFT accelerator is carried out via a series of design and simulation works. The paper focuses the influence of different magnetic field configurations on the stray electrons and the character of the ejecting electrons.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Journal of Powder Materials
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• v.31 no.3
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• pp.220-225
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• 2024

Molten salt reactors represent a promising advancement in nuclear technology due to their potential for enhanced safety, higher efficiency, and reduced nuclear waste. However, the development of structural materials that can survive under severe corrosion environments is crucial. In the present work, pure Ni was deposited on the surface of 316H stainless steel using a directed energy deposition (DED) process. This study aimed to fabricate pure Ni alloy layers on an STS316H alloy substrate. It was observed that low laser power during the deposition of pure Ni on the STS316H substrate could induce stacking defects such as surface irregularities and internal voids, which were confirmed through photographic and SEM analyses. Additionally, the diffusion of Fe and Cr elements from the STS316H substrate into the Ni layers was observed to decrease with increasing Ni deposition height. Analysis of the composition of Cr and Fe components within the Ni deposition structures allows for the prediction of properties such as the corrosion resistance of Ni.

• Korean Journal of Materials Research
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• v.9 no.11
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• pp.1137-1143
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• 1999

Microstructure and mechanical properties of Mod.9Cr-1Mo and W added 9Cr-0.5Mo2W steels were investigated for liquid metal reactor (LMR) heat exchange tube. The tempering temperatures at which cell structure was formed were $700^{\circ}C$ for Mod.9Cr-1Mo steel and $750^{\circ}C$ for W added 9Cr0.5Mo-2W steel. indicating the recovery of dislocation was delayed by the addition of W. 9Cr-0.5Mo-2W steel had the same kinds of precipitates with Mod.9Cr-1Mo steel, but the W was included in the precipitates in 9Cr-0.5Mo-2W steel. Micro-hardness and ultimate tensile strength of 9Cr-0.5Mo-2W steel were higher than those of Mod.9Cr-1Mo steel. The impact property of Mod.9Cr-1Mo steel was superior to that of 9Cr-0.5Mo-2W steel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.33 no.1
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• pp.63-72
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• 2020

The purpose of this study is to investigate the effects of the application of various numerical models and frequency contents of earthquakes on the performances of the reactor containment building (RCB) in a nuclear power plant (NPP) equipped with an advanced power reactor 1400. Two kinds of numerical models are developed to perform time-history analyses: a lumped-mass stick model (LMSM) and a full three-dimensional finite element model (3D FEM). The LMSM is constructed in SAP2000 using conventional beam elements with concentrated masses, whereas the 3D FEM is built in ANSYS using solid elements. Two groups of ground motions considering low- and high-frequency contents are applied in time-history analyses. The low-frequency motions are created by matching their response spectra with the Nuclear Regulatory Commission 1.60 design spectrum, whereas the high-frequency motions are artificially generated with a high-frequency range from 10Hz to 100Hz. Seismic responses are measured in terms of floor response spectra (FRS) at the various elevations of the RCB. The numerical results show that the FRS of the structure under low-frequency motions for two numerical models are highly matched. However, under high-frequency motions, the FRS obtained by the LMSM at a high natural frequency range are significantly different from those of the 3D FEM, and the largest difference is found at the lower elevation of the RCB. By assuming that the 3D FEM approximates responses of the structure accurately, it can be concluded that the LMSM produces a moderate discrepancy at the high-frequency range of the FRS of the RCB.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.35-46
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• 1997

A simple and precise method of $^{14}C$ was developed to analyze $^{14}C$ in the environment samples using a commercially available $^{14}CO_2$ absorbent and a liquid scintillation counter. An air sampler and a combustion system were developed to collect HTO and $^{14}CO_2$ in the air and the biological samples simultaneously. The collection yield of $^{14}CO_2$ by the air sampler was in the range of 73-89% . The yield of the combustion system was 97%. In preparing samples for counting, the optimum ratio of $CO_2$ absorbent to the scintillator for mixing was 1:1. No variation of the specific activity of $^{14}C$ in the counting sample was observed up to 70 days after preparation of the samples. The detection limit for$^{14}C$ was 0.025 Bq/gC, which is the level applicable to the natural level of $^{14}C$. The analytical result of $^{14}C$ obtained by the present method were within ${\pm}6%$ of the relative error from the one by the benzene synthesis. The specific activity of $^{14}C$ in the air collected at Taejon during the period of October 1996 ranged from 0.26 to 0.27 Bq/gC. The specific activity of $^{14}C$ in the air collected at 1km from the Wolsong nuclear power plant a 679 MWe PHWR, was $0.54{\pm}0.03$ Bq/gC. The ranges of specific activities of $^{14}C$ in the pine needles and the vegetations from the areas around the Wolsong nuclear power plant were 0.56-0.67 Bq/gC and 0.23-1.41 Bq/gC, respectively.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.12
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• pp.3748-3754
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• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.