• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.511-521
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• 2022

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y^-1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.1-8
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• 2023

The major concern in the deep geological disposal of spent nuclear fuels include sulfide-induced corrosion and stress corrosion cracking of copper canisters. Sulfur diffusion into copper canisters may induce copper embrittlement by causing Cu₂S particle formation along grain boundaries; these sulfide particles can act as crack initiation sites and eventually cause embrittlement. To prevent the formation of Cu₂S along grain boundaries and sulfur-induced copper embrittlement, copper alloys are designed in this study. Alloying elements that can act as chemical anchors to suppress sulfur diffusion and the formation of Cu₂S along grain boundaries are investigated based on the understanding of the microscopic mechanism of sulfur diffusion and Cu₂S precipitation along grain boundaries. Copper alloy ingots are experimentally manufactured to validate the alloying elements. Microstructural analysis using scanning electron microscopy with energy dispersive spectroscopy demonstrates that Cu₂S particles are not formed at grain boundaries but randomly distributed within grains in all the vacuum arc-melted Cu alloys (Cu-Si, Cu-Ag, and Cu-Zr). Further studies will be conducted to evaluate the mechanical and corrosion properties of the developed Cu alloys.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1442-1450
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• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Journal of the Korean Ceramic Society
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• v.45 no.4
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• pp.245-249
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• 2008

The SiC and ZrC are critical and essential materials in TRISO coated fuel particles since they act as protective layers against diffusion of metallic and gaseous fission products and provides mechanical strength for the fuel particle. However, SiC and ZrC have critical disadvantage that SiC loses chemical integrity by thermal dissociation at high temperature and mechanical properties of ZrC are weaker than SiC. In order to complement these problems, we made new combinations of the coating layers that the ZrC layers composed of SiC. In this study, after Silicon carbide(SiC) were chemically vapor deposited on graphite substrate, Zirconium carbide(ZrC) were deposited on SiC/graphite substrate by using Zr reaction technology with Zr sponge materials. The different morphologies of sub-deposited SiC layers were correlated with microstructure, chemical composition and mechanical properties of deposited ZrC films. Relationships between deposition pressure and microstructure of deposited ZrC films were discussed. The deposited ZrC films on SiC of faceted structure with smaller grain size has better mechanical properties than deposited ZrC on another structure due to surface growth trend and microstructure of sub-deposited layer.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Geomechanics and Engineering
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• v.31 no.5
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• pp.461-476
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• 2022

Research studies on the scale effect on triaxial strength of intact rocks are scarce, being more common those in uniaxial strength. In this paper, the authors present and briefly interpret the peak and residual strength trends on a series of triaxial tests on different size specimens (30 mm to 84 mm diameter) of an intact granitic rock at confinements ranging from 0 to 15 MPa. Peak strength tends to grow from smaller to standard-size samples (54 mm) and then diminishes for larger values at low confinement. However, a slight change in strength is observed at higher confinements. Residual strength is observed to be much less size-dependent. Additionally, this study introduces preliminary modelling approaches of these laboratory observations with the help of three-dimensional particle flow code (PFC3D) simulations based on bonded particle models (BPM). Based on previous studies, two modelling approaches have been followed. In the first one, the maximum and minimum particle diameter (D_max and D_min) are kept constant irrespective of the sample size, whereas in the second one, the resolution (number of particles within the sample or ϕ_v) was kept constant. Neither of these approaches properly represent the observations in actual laboratory tests, even if both of them show some interesting capabilities reported in this document. Eventually, some suggestions are provided to proceed towards improving modelling approaches to represent observed scale effects.

• 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.

• Applied Chemistry for Engineering
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• v.8 no.3
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• pp.430-437
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• 1997

Uranyl-peroxide compound was prepared by the reaction of excess hydrogen peroxide solution and trace uranium in filtrate from nuclear fuel conversion plant. The $CO_3{^{2-}}$ in filtrate was removed first by heating more than $98^{\circ}C$, because uranyl-peroxide compound could not be precipitated by $CO_3{^{2-}}$ remaining in filtrate. The optimum condition for uranyl-peroxide compound was ageing for 1 hr after controling the pH with $NH_3$ gas and adding the excess $H_2O_2$ of 10ml/lit.-filtrate. Uranium concentration in the filtrate was appeared to 3 ppm after the precipitation of uranyl-peroxide compound, and the chemical composition of this compound was analyzed to $UO_4{\cdot}2NH_4F$ with FT-IR, X-ray diffractometry, TG and chemical analysis. Also, this fine particle, about $1{\sim}2{\mu}m$, could be grown up to $4{\mu}m$ at pH 9.5 and $60^{\circ}C$. The separation efficiency of precipitate from mother liquor was increased with increase of pH and reaction temperature. Otherwise, the crystal form of this particle showed octahedral by SEM and XRD, and $U_3O_8$ powder was obtained by thermal decomposition at $650^{\circ}C$ in air atmosphere.

• Korean Chemical Engineering Research
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• v.46 no.1
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• pp.56-62
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• 2008

Characteristics of heat transfer were investigated in a three-phase swirling fluidized bed whose diameter was 0.102 m and 2.5 m in height. Effects of gas and liquid velocities, particle size and liquid swirling ratio ($R_S$) on the immersed heater-to-bed overall heat transfer coefficient were examined. The heat transfer characteristics between the immersed heater and the bed was well analyzed by means of phase space portraits and Kolmogorov entropy(K) of the time series of temperature difference fluctuations. The phase space portraits of temperature difference fluctuations became stable and periodic and the value of Kolmogorov entropy tended to decrease with increasing the value of liquid swirling ratio from 0.1 to 0.4. The value of Kolmogorov entropy exhibited its minimum with increasing liquid swirling ratio. The value of overall heat transfer coefficient (h) showed its maximum with the variation of liquid velocity, bed porosity or liquid swirling ratio, but it increased with increasing gas velocity and particle size. The value of K exhibited its maximum at the liquid velocity at which the h value attained its maximum. The overall heat transfer coefficient and Kolmogorov entropy were well correlated in terms of dimensionless groups and operating variables.