• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.608-616
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• 2015

The migration of silver (Ag) in silicon carbide (SiC) and $^{110m}Ag$ through SiC of irradiated tristructural isotropic (TRISO) fuel has been studied for the past three to four decades. However, there is no satisfactory explanation for the transport mechanism of Ag in SiC. In this work, the diffusion coefficients of Ag measured and/or estimated in previous studies were reviewed, and then pre-exponential factors and activation energies from the previous experiments were evaluated using Arrhenius equation. The activation energy is $247.4kJ{\cdot}mol^{-1}$ from Ag paste experiments between two SiC layers produced using fluidized-bed chemical vapor deposition (FBCVD), $125.3kJ{\cdot}mol^{-1}$ from integral release experiments (annealing of irradiated TRISO fuel), $121.8kJ{\cdot}mol^{-1}$ from fractional Ag release during irradiation of TRISO fuel in high flux reactor (HFR), and $274.8kJ{\cdot}mol^{-1}$ from Ag ion implantation experiments, respectively. The activation energy from ion implantation experiments is greater than that from Ag paste, fractional release and integral release, and the activation energy from Ag paste experiments is approximately two times greater than that from integral release experiments and fractional Ag release during the irradiation of TRISO fuel in HFR. The pre-exponential factors are also very different depending on the experimental methods and estimation. From a comparison of the pre-exponential factors and activation energies, it can be analogized that the diffusion mechanism of Ag using ion implantation experiment is different from other experiments, such as a Ag paste experiment, integral release experiments, and heating experiments after irradiating TRISO fuel in HFR. However, the results of this work do not support the long held assumption that Ag release from FBCVD-SiC, used for the coating layer in TRISO fuel, is dominated by grain boundary diffusion. In order to understand in detail the transport mechanism of Ag through the coating layer, FBCVD-SiC in TRISO fuel, a microstructural change caused by neutron irradiation during operation has to be fully considered.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 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.

• Journal of the Korean Chemical Society
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• v.15 no.4
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• pp.191-197
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• 1971

Ruthenium content in highly purified palladium metal (99.9%) was determined by counting $^{105}Rh$ nuclide which was produced by $^{104}Ru(n,{\gamma};{\beta}^-)^{105}Rh$ nuclear reaction. Palladium sample and ruthenium standard were irradiated by neutron with the Pneumatic Transfer System of TRIGA MARK II reactor. Palladium and ruthenium were dissolved by treating with aqua-regia and by fusing with sodium peroxide flux respectively. $^{105}Rh$ was separated through anion and cation exchange resin columns. The ruthenium content was determined by comparing the $^{105}Rh$ activities, obtained from the palladium sample, with that from pure ruthenium standard. The detection limit of ruthenium by the present method is about 1 ppm of ruthenium in 10 mg of palladium, which is approximately 40 times more sensitive than that of the conventional radioactivation method which employs $^{102}Ru(n,{\gamma})^{103}Ru$ nuclear reaction.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Journal of Korean Society of Water and Wastewater
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• v.25 no.5
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• pp.681-689
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• 2011

Submerged membrane bio-reactor (SMBR) has several advantages such as high MLSS, long SRT, and low F/M ratio at wastewater treatment. So, this has widely applied over the world and many studies have been conducted. However, membrane fouling remains an inevitable problem. This study was investigated using bench-scale SMBR with three poeration modes. Raw waters were prepared by addition of starch, acetic and fibric acid to recovery water of zeolite. The efficiency of nitrification and COD were very stable as about 95% and 80%, respectively. And critical flux was 128.8L/$m^{2}$/hr. The result of biodegradability test was following values at the each mode : Ss+Xs/$C_{T}$=81.7%, 35.1% and 45.3%, $X_{I}+S_{I}/C_{T}=18.3%$, 64.9% and 54.7%. When particulate matters such as $X_{I}$ and $X_{S}$ in influent are increased, membrane fouling will take place more and more. A relative ratio of filtration resistance to the fouling occurred by the cake layer was increased when increased the portion of $X_{I}$ and polysaccharide. It was thought that the formation of cake layer was promoted due to bond between $X_{I}$ and vicid material s generated from the polysaccharide.

• Journal of Korean Society of Water and Wastewater
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• v.13 no.1
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• pp.61-71
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• 1999

For a high-rate fermentation and recovery of organic acid, we have developed a new organic acid fermentation reactor with membrane filter, which is the most important part in the new advanced wastewater treatment system. The recovered organic acid is to be reused as an organic carbon source at denitrification process. Some experiments were conducted to compare the performance of acid fermentation at different SRTs, such as 5, 10, and 20 days. The total organic acid concentration produced during the runs was in the range of 2,100-2,900 (mgC/L). The conversion efficiency from substrate to organic acid reached to from 43% to 59%. The recovery rate of organic acid from substrate based on TOC was from 26% to 53%. Regardless of operational conditions, it has been able to maintain the membrane flux constantly, in the range of 0.4-0.46 ($m^3/m^2/day$). The transmembrane pressure drop was 0.2-0.3 (kg/cm) for 100 day's operation. The result of simulation is as follows. Organic removal efficiency of the new advanced treatment system is 95%. 73% of Nitrogen is removed. The removal efficiency of Phosphorus is 93%. By coqulation, soluble phosphorus is able to remove from the water treatment lines, which is impossible at conventional activated sludge system. The unit construction cost is 65000 (yen/m3) and it was 1.4 times than that of the standard activated sludge system. The unit operation cast is 7.7 ($yen/m^3/day$) and it was 1.3 times than that of the standard activated sludge system.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.608-616
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• 2022

Power peaking factors (PPF) is an important parameter for safe and efficient reactor operation. There are several methods to calculate the PPF at TRIGA research reactors such as MCNP and TRIGLAV codes. However, these methods are time-consuming and required high specifications of a computer system. To overcome these limitations, artificial intelligence was introduced for parameter prediction. Previous studies applied the neural network method to predict the PPF, but the publications using the ANFIS method are not well developed yet. In this paper, the prediction of PPF using the ANFIS was conducted. Two input variables, control rod position, and neutron flux were collected while the PPF was calculated using TRIGLAV code as the data output. These input-output datasets were used for ANFIS model generation, training, and testing. In this study, four ANFIS model with two types of input space partitioning methods shows good predictive performances with R² values in the range of 96%-97%, reveals the strong relationship between the predicted and actual PPF values. The RMSE calculated also near zero. From this statistical analysis, it is proven that the ANFIS could predict the PPF accurately and can be used as an alternative method to develop a real-time monitoring system at TRIGA research reactors.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4335-4354
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• 2024

Some core designs integrate high-enriched fuel and moderator materials to enhance neutron utilization. This combination results in a broad spectrum within the system, posing challenges in resonance calculation. This paper introduces a general framework to realize resonance self-shielding treatment in broad-spectrum fuel lattice problems. The framework consists of three components. First, a new energy group structure is devised to support resonance calculation in the entire energy range and capture spectral transition and thermalization effects during eigenvalue calculation. Second, the subgroup method based on narrow approximation is selected as a universal method to perform resonance calculation. Finally, transport equations for each fissionable region are solved for neutron flux to collapse the fission spectrum. The proposed method is verified against fast, intermediate, and thermal spectrum pin cell problems and an assembly problem featuring a fast-thermal coupled spectrum. Numerical results affirm the accuracy of the proposed method in handling these scenarios, with eigenvalue errors below 154 pcm for pin cell problems and 106 pcm for the assembly problem. The verification results revealed that the proposed method enables accurate resonance self-shielding treatment for broad-spectrum problems.

• Journal of Powder Materials
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• v.15 no.3
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• pp.182-187
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• 2008

Carbon was known to be one of effective additives which can improve the flux pinning of $MgB_2$ at high magnetic fields. In this study, glycerin $(C_3H_8O_3)$ was selected as a chemical carbon source for the improvement of critical current density of $MgB_2$. In order to replace some of boron atoms by carbon atoms, the boron powder was heat-treated with liquid glycerin. The glycerin-treated boron powder was mixed with an appropriate amount of magnesium powder to $MgB_2$ composition and the powder pallets were heat treated at $650^{\circ}C\;and\;900^{\circ}C$ for 30 min in a flowing argon gas. It was found that the superconducting transition temperature $(T_c)$ of $Mg(B_{1-x}C_x)_2$ prepared using glycerin-treated boron powder was 36.6 K, which is slightly smaller than $T_c$(37.1 K) of undoped $MgB_2$. The critical current density $(J_c)$ of $Mg(B_{1-x}C_x)_2$ was higher than that of undoped $MgB_2$ and the $T_c$ improvement effect was more remarkable at higher magnetic fields. The $T_c$, decrease and $J_c$ increase associated with the glycerin treatment for boron powder was explained in terms of the carbon substitution to boron site.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.842-848
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• 2022

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 ℃. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner subchannel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.