• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Bulletin of the Korean Chemical Society
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• v.35 no.5
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• pp.1397-1402
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• 2014

The complete active space self-consist field method followed by the internally contracted multireference configuration interaction method has been used to compute the potential energy curves of $X^2\prod_g$, $a^4\prod_u$, $A^2\prod_u$, $b^4\sum_{g}^{-}$, and $B^2\sum_{g}^{-}$ states of $S{_2}^+$ cation with large correlation-consistent basis sets. Utilizing the potential energy curves computed with different basis sets, the spectroscopic parameters of these states were evaluated. Finally, the transition dipole moment and the Franck-Condon factors of the transition from $A^2\prod_u$ to $X^2\prod_g$ were evaluated. The radiative lifetime of $A^2\prod_u$ is calculated to be 887 ns, which is in good agreement with experimental value of $805{\pm}10$ ns.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.83-92
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• 2013

A novel methodology to evaluate remote dismantling equipment for a reactor pressure vessel (RPV) in a decommissioning project is presented in this paper. The remote dismantling equipment, mainly composed of cutting tools and positioning equipment, is absolutely required to cut and handle highly radioactive and large components in nuclear power plants (NPPs); this equipment has a great effect on the overall success of the decommissioning project. Conventional evaluation methods have only focused on cutting technologies or positioning equipment, although remote dismantling equipment cannot achieve its goal without organic interaction between the cutting tools and the positioning equipment. In this paper, the cutting tools and the positioning equipment are evaluated by performance parameters according to their original characteristics, the relationship between the two systems, and common factors. Finally, the remote dismantling equipment used in recent decommissioning projects has been evaluated based on the proposed methodology. The results of this paper are expected to be useful for future decommissioning projects.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.2025-2037
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• 2021

The influence of lead cations on natural quartz (QZ) from Egypt as a glass shielding material for the composition with nominal formula (10Na₂O - (90 - x) QZ - xPbO (where x = 30, 35, 40, 45 and 50 mol %)) was examined. The studied samples are synthesized via the melt quenching method at 1050 ℃. The X-ray diffraction XRD patterns were confirmed the glass nature for studied samples. Moreover, the optical properties, and the transparency for all compositions were examined by UV-Vis spectroscopy. Also, the major elemental composition of the natural quartz were estimated via the X-ray fluorescence (XRF) technique. Further, the density and molar volume were determined. Furthermore, the nuclear shielding parameters such as, mass attenuation coefficient, effective atomic number, electronic density, the total atomic, and electronic cross sections as well as the mean free path, and the half value layer with different gamma ray energies (81 keV-1407 keV) were calculated. Besides, the results showed that the shielding behavior towards the gamma ray radiation for all glass samples was increased as the increment in PbO concentration in the glass system.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.533-544
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• 2002

The neutron induced nuclear data for Mo-95, Tc-99, Ru-101 and Rh-103 was calculated and evaluated in the fast energy region. The energy dependent optical model potential parameters were extracted based on the recent experimental data and applied up to 20 MeV. The s-wave strength function was calculated from the parameters. Spherical optical model, statistical model in equilibrium energy, multistep direct and multistep compound model in pre-equilibrium energy and direct capture model were used in the calculation. The theoretically calculated cross sections were compared with the experimental data and the evaluated files The model- calculated total and capture cross sections were in good agreement with the reference experimental data. The direct capture contribution improved the capture cross sections in pre- equilibrium region. The evaluated cross section results were compiled to ENDF-6 format and will improve the ENDF/B-Vl.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.546-554
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• 2023

The cyclical behavior of Alloy 617 was examined at 25 ℃ and high temperatures of 800, 850, 900, and 950 ℃ in air to obtain its fatigue life curves. The specimens tested at 25, 800, and 850 ℃ cyclically hardened, whereas those tested above 900 ℃ cyclically softened from the first cycle, that is, their fatigue life was reduced at high temperatures owing to loss of strength. Parameters of the typical Coffin-Manson-Basquin relationship were determined for each test temperature. Interestingly, no significant difference in fatigue life was observed for the specimens tested in the range of 800-950 ℃. Owing to the similarity in fatigue life, we determined fatigue strength and fatigue ductility exponents that could be applied for this temperature range. The parameters obtained were close to the universal slopes, although the fatigue ductility exponent was slightly different. The proposed fatigue life curves were compared with those presented in ASME code.

• Macromolecular Research
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• v.10 no.5
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• pp.291-295
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• 2002

Atomic force microscopy(AFM) was used to study the polyethylene(PE) surfaces grafted and immobilized with acrylic acid by Ar plasma treatment. The topographical images and parameters including RMS roughness and Rp-v value provided an appropriate means to characterize the surfaces. The plasma grafting and immobilization method were a useful tool for the preparation of surfaces with carboxyl group. However, the plasma immobilization method turned out to have a limitation to use as a means of preparation of PE surface with specific functionalities, due to ablation effect during the Ar plasma treatment process.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.385-392
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• 1997

The shape of TL glow curve is a useful indicator for assurance of correct reading of the personal dosimeter. Since the reading procedure of TLD is irreversible, however, an analytic remedy should be considered to procure reliable dosimetric information for the readings with irregular glow con shape. In this study, kinetic trapping parameters of CaSO$_4$ : Dy Teflon personal dosimeter(Teledyne PB-6A) were analyzed by Halperin and Braner's model for general-order kinetics. From these kinetic tapping parameters, we also developed a simple procedure to retrieve the dosimetric information from abnormally distorted glow curves. The computerized glow curve deconvolution(CGCD) fitting of the reference glow curve with kinetic parameters from this study yields relative errors of about 5% from the expected integral. It was also found that the glow curve remedial procedure developed could retrieve the distorted TL glow curves within ewer ranges of 1575. With the glow curve retrieval techniques, doses incurred by gamma radiation can now be successfully re-constructed for the CaSO$_4$ : Dy Teflon dosimeter resulting abnormal glow curves.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.165-172
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• 2016

An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.