• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.277-286
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• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

• Analytical Science and Technology
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• v.17 no.4
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• pp.355-362
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• 2004

Fission products and Inter reaction layer of $U_3Si-Al$ dispersion fuel, irradiated in HANARO research reactor with 121 kW/m of maximum liner power and 63 at% of average burn-up, was characterization by EPMA (Electron Probe Micro Analyzer). The fuel punching system developed by Irradiated Materials Examination Facility (IMEF) has used to make these samples for the EPMA. With this system a very small and thin specimen which is 1.57 mm in diameter and 2 mm in thickness respectively has been fabricated to protect the EPMA operator from high radioactive fuel and to mini-mize the equivalent dose rate less than 150 mSv/h. EPMA was performed to observe layers of sectional, Inter-reaction and oxide with specimens of cutting and polished. Stoichiometry in the Inter-reaction layer with $16{\mu}m$ of thickness was $U_{2.84}$ Si $Al_{14}$ with calibration of $UO_2$ and $U_{3.24}$ Si $Al_{14.1}$ with calibration of standard specimen. metallic precipitates in this layer were not observed using fission products examination.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.55-63
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• 2016

Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy ($E_e$) and source multiplication coefficient ($k_s$), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient ($k_{eff}$), net neutron multiplication (M), neutron yield ($Y_{n/e}$), energy constant gain ($G_0$), energy gain (G), importance of neutron source (${\varphi}^*$), axial and radial distributions of neutron flux, and power peaking factor ($P_{max}/P_{ave}$) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current ($I_e$) have been decreased in the highest case of $k_s$, but G and ${\varphi}^*$ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing $E_e$ from 100 MeV up to 1 GeV, $Y_{n/e}$ and G improved by 91.09% and 10.21%, and $I_e$ and $P_{acc}$ decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np-Pu assemblies on the periphery allows for a consistent $k_{eff}$ because the Np-Pu assemblies experience less burn-up.

• Analytical Science and Technology
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• v.13 no.6
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• pp.751-758
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• 2000

A study was carried out on the dissolution of spent PWR fuels and performed on the fuels and the separation of iodide for the quantitative analysis using SIMFUEL which has chemical composition of a simulated spent PWR fuel (burn-up; 35,000 MWd/MTU and cooling time; 10 years). To dissolve the SIMFUEL effectively and to minimize the formation of volatile iodine through dissolution process, the optimum ratio of mixed acid ($HNO_3/HCl$ 80: 20 mol%) was established and ozone gas was purged. In the separation step of iodine with $CCl_4$, $NH_2OH{\cdot}HCl$ was used for reducing ${IO_3}^-$ to $I_2$.The optimum acidity of the dissolved solution and the added of $NH_2OH{\cdot}HCl$ were 2.5 M and more than $1.5{\times}10^{-3}mole$, respectively. The recovery of iodide by ion chromatography was $82.8{\pm}4.1%$ and the total yield was corrected by gamma spectrometery using $^{131}I$ as a tracer.

• Transactions of the Korean Society of Automotive Engineers
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• v.14 no.3
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• pp.170-177
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• 2006

Hydrogen can extend the lean misfire limit to a large extent when it is mixed with conventional fuels for a spark ignition engine. In this study, hydrogen-enriched gaseous fuels by reforming process were simulated according to their proportions of $H_2$, CO, $CO_2$ and $N_2$ gases. Pure hydrogen and two different hydrogen-enriched gaseous mixtures(A-, B-composition) were tested for their basic effects on the engine performances and emissions in a single cylinder research engine. A- and B-composition showed different results from 100% $H_2$ addition because air/fuel mixtures were more diluted by their additions. Even though the energy fraction of reformed gases was increased, combustion stabilities and lean misfire limits were not sensitively improved. It means that combustion augmentation by $H_2$ addition was offset by the charge dilution of $N_2$ and $CO_2$. In addition, the low flammability of CO gas deteriorated thermal efficiencies. CO emission was drastically increased with B-composition which included higher CO component. However, $NO_x$ was reduced as energy fraction($X_e$) rised except for the case of 100% $H_2$ addition at $\lambda=1.2$ and was, for A-composition, lowered to a factor of ten when compared with that of $H_2$ addition. HC emissions were largely influenced by $COV_{imep}$ due to misfire and partial burns.

• Journal of architectural history
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• v.27 no.3
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• pp.15-26
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• 2018

As a result of reviewing various documents and existing researches, since the late Goryeo period, the most active period in the Ondol(溫突, Korean floor heating system) facilities is the 17th century. The phenomenological reason was recovering the buildings destroyed by the Japanese Invasion of Korea in 1592(壬辰倭亂) & the Manchu War of 1636(丙子胡亂), but the underlying cause was an abnormal climate in which a pair of summer and winter cold continued. In the 17th century, as the Ondol facilities grew rapidly without distinction between regions and classes, the supply and demand of fuel caused economic and natural environmental problems. And a negative and positive view on Ondol was suggested. Since the middle of the 18th century, when the demand and supply of Ondol reached its peak, which could no longer increase, a new awareness of Ondol began to grow. The room was called the Panbang(板房) and the Ondol, depending on the material that made up the floor. It was considered natural to have the Ondol from this time on. The Incan(因間) and Jo(竈) that were made to burn were started to be recorded as a kitchen, regardless of size and function. Changes in social awareness of Ondol have led to concerns about heating efficiency. A variety of architectural explorations were conducted. Such a search was later realized in concrete architectural form. There is a double Ondoll structure, and the column spacing is reduced compared to the previous one. The heat buffer space is formed around the Ondol room, and the double window can control the light and the air going in and out.

• Transactions of the Korean hydrogen and new energy society
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• v.24 no.1
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• pp.29-35
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• 2013

A dense mixed ionic and electronic conducting ceramic membrane is one of the most promising materials because it can be used for separation of oxygen from the mixture gas. The $ABO_3$ perovskite structure shows high chemical stability at high temperatures under reduction and oxidation atmospheres. $BaCo_{1-x-y}Fe_xZr_yO_{3-{\delta}}$ (BCFZ) was well-known material as high mechanical strength, low thermal conductivity and stability in the high valence state. Glycine Nitrate Process (GNP) is rapid and effective method for powder synthesis using glycine as a fuel and show higher product crystallinity compared to solid state reaction and citrate-EDTA method. BCFZ was fabricated by modified glycine nitrate process. In order to control the burn-up reaction, $NH_4NO_3$ was used as extra nitrate. According to X-Ray Diffraction (XRD) results, BCFZ was single phase regardless of Zr dopants from y=0.1 to 0.3 on B sites. The green compacts were sintered at $1200^{\circ}C$ for 2 hours. Oxygen permeability, methane partial oxidation rate and hydrogen production ability of the membranes were characterized by using Micro Gas Chromatography (Micro GC) under various condition. The high oxygen permeation flux of BCFZ 1-451 was about $1ml{\cdot}cm^{-2}s^{-1}$. Using the humidified Argon gas, BCFZ 1-433 produced hydrogen about $1ml{\cdot}cm^{-2}s^{-1}$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.21-34
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• 2015

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.