• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.283-288
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• 1996

Direct safety injection into the reactor vessel downcomer annulus(DVI) is a fundamental feature of the KNGR(Korean Next Generation Reactor) four-train safety injection system. The numerical analysis of thermal mixing of ECC(Emergency Core Cooling) water through DVI with the water in the RVDC(Reactor Vessel Downcomer) annulus has been performed, in order to study the impact of nozzle location on the pressurized thermal shock and safety analysis. The results of this study show that the thermal mixing due to the natural circulation induced by the limiting accident conditions is sufficient to prevent temperature in the RVDC from dropping to the level of concern for PTS. When the DVI nozzle is located right above the cold leg, the temperature distribution at the outlet of flow field is most uniform. The tool used for numerical analysis is CFDS-FLOW3D.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.862-868
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• 1998

Korean Next Generation Reactor(KNGR) is a standardized evolutionary Advanced Light Water Reactor design under development Korea Power Company(KEPCO). It incorporates design enhncements such as active and passive advanced design features(ADFs) to increase the plant safety. A Preliminary level 1 Probabilistic Safety Assessment(PSA) has been performed for KNGR to examine the effect of these safety features. The preliminary PSA result shows that it meets the KNGR safety goal on core damage frequency(CDF). The result of this safety assessment shows that the four-train safety systems, and the ADFs such as Passive Secondary Cooling System (PSCS) contributes greatly to the reduction the CDF. Furthermore, several design changes are made or proposed for detailed review based on the PSA insights.

• Proceedings of the KIEE Conference
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• 1994.07b
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• pp.1431-1434
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• 1994

The annular induction electromagnetic pump with maximum flowrate of $60{\ell}/min$ for the sodium coolant system of liquid metal fast breeder reacters has been designed using the equivalent circuit method. The final optimum values of geometrical and electromagnetic parameters were obtained for an annular induction pump from the relation of the electrical variables giving the developing force to the fluid and the pressure drops between both sides of the pump. The physical properties of the core, coil condoctor materials in the high temperature and pump cooling systems under operation have been taken into account in the design of the pump. The structural material were also selected considering the reaction with sodium and the magnetic field distortion.

• Journal of the Semiconductor & Display Technology
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• v.1 no.1
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• pp.9-13
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• 2002

A polyimide-type electrostatic chuck (ESC) was fabricated for the application of holding 8-inch silicon wafers in the oxide etching equipment. For the fabrication of the unipolar ESC, core technologies such as coating of polyimide films and anodizing treatment of aluminum surface were developed. The polyimide films were prepared on top of thin coated copper substrates for the good electrical contacts, and the helium gas cooling technique was used for the temperature uniformity of the silicon wafers. The ESC was essentially working with an unipolar operation, which was easier to fabricate and operate compared to a bipolar operation. The chucking force of the ESC has been measured to be about 580 gf when the applied voltage was 1.5 kV, which was considered to be enough force to hold wafers during the dry etching processing. The employment of the ESC in etcher system could make 8% enhancement of the wafer processing yield.

• Journal of the Korean Society of Visualization
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• v.20 no.2
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• pp.28-37
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• 2022

Flow characteristics of jet impingement over concave hemispherical surface with effusion cooling holes is relatively more complex than that of a flat surface, so the experimental validation for computational fluid dynamics (CFD) results is important. In this study, experimental results were compared with simulation results obtained by assuming different turbulence models. The vortex was observed in the region between the central jets where the recirculation structure appeared. The different patterns of vorticity distributions were observed for each turbulence models due to different interaction of the injected jet flow. Among them, the transition k-kl-ω model predicted similarly not only the jet potential core region with higher velocity, but also the recirculation region between the central jets. From the validation, it may be helpful to accurately predict heat and mass transfer in jet impingement/effusion hole system.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.2
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• pp.213-218
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• 2010

A cooling system utilizing liquid helium to chill the cryopanel (800 mm $\times$ 700 mm dimensions) down to 4.2 K was designed, implemented, and tested to verify the role of the cryopanel as a heat sink for the payload of a spacecraft inside the large thermal vacuum chamber (effective dimensions : 8 m ($\Phi$) $\times$ 10 m (L)) of KARI (Korea Aerospace Research Institute). Two LHe (Liquid Helium) Dewars, one for the main supply and the other for refilling, were used to supply liquid helium or cold helium gas into this cryopanel, and flow control for the target temperature of the cryopanel within requirements was done through fine adjustment of the pressure inside the LHe Dewars. The return helium gas from the cryopanel was reused as a thermal barrier to minimize the heat influx on the core liquid helium supply pipe. The test verified a cooling time of around three hours from the ambient temperature to 40 K (combined standard uncertainty of 194 mK), the capacity for maintaining the cryopanel at intermediate temperatures, and a 1 K uniformity over the entire cryopanel surface at around 40 K with 20 W cooling power.

• Proceedings of the KIEE Conference
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• 2002.07b
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• pp.943-945
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• 2002

This paper presents conceptual design and 3-D electromagnetic analysis of 1MVA transformer with BSCCO-2223 High Tc Superconducting (HTS) tapes. The rated voltages of each sides of the transformer are 22.9 kV and 6.6 kV, and double pancake windings were adopted. High voltage and Low voltage sides were composed of several double pancake windings. Four HTS tapes were wound in parallel for the windings of low voltage side and were transposed in order to distribute the currents equally in each conductor. The transformer core was designed as a shell type core made of laminated silicon steel plates and the core is separated with the windings by a cryostat with Fiberglass Reinforced Plastics(FRP). A sub-cooling system using $LN_2$ were designed to maintain the coolant temperature of 65K. Finally perpendicular components of magnetic field applied to tapes were calculated 0.24T in the rated operation using 3-D analysis. A real 1MVA HTS transformer will be manufactured in near future based on the design parameters presented in this paper.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.