• Nuclear Engineering and Technology
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• 제47권1호
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• pp.33-46
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• 2015

During the past 10 years, the Korea Atomic Energy Research Institute (KAERI) has performed a study to control hydrogen gas in the containment of the nuclear power plants. Before the Fukushima accident, analytical activities for gas distribution analysis in experiments and plants were primarily conducted using a multidimensional code: the GASFLOW. After the Fukushima accident, the COM3D code, which can simulate a multidimensional hydrogen explosion, was introduced in 2013 to complete the multidimensional hydrogen analysis system. The code validation efforts of the multidimensional codes of the GASFLOW and the COM3D have continued to increase confidence in the use of codes using several international experimental data. The OpenFOAM has been preliminarily evaluated for APR1400 containment, based on experience from coded validation and the analysis of hydrogen distribution and explosion using the multidimensional codes, the GASFLOW and the COM3D. Hydrogen safety in nuclear power has become a much more important issue after the Fukushima event in which hydrogen explosions occurred. The KAERI is preparing a large-scale test that can be used to validate the performance of domestic passive autocatalytic recombiners (PARs) and can provide data for the validation of the severe accident code being developed in Korea.

• 산업경영시스템학회지
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• 제34권4호
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• pp.49-56
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• 2011

The Project of Jordan Research and Training Reactor (JRTR) officially launched in Aug. 2010. JRTR is the first made-in-Korea nuclear system to be built abroad by year 2015, and Korea Atomic Energy Research Institute (KAERI) is responsible for the design of major systems including the reactor core. While the PDCS (Project Document Control System) being operated by EPC company controls all the documents of the whole Project, KAERI is supposed to have its own system for KAERI documents. Meeting such a need; KAERI has implemented a document control for the JRTR Project into already existing ANSIM (KAERI Advanced Nuclear Safety Information Management) system. The documents of JRTR project to be controlled are defined in the PPM (Project Procedures Manual), QAP (Quality Assurance Procedure) and PEP (Project Execution Program). The ANSIM consists of the document management holder, document container holder and organization management holder. The document management holder, which is the most important part of ANSIM-JRTR, consists of the DDA (Document Distribution for Agreement), IOC (Inter-office Correspondence), PM Memo. (Project Manager Memorandum) and cover sheets of design documents. Other materials such as meeting minutes, sub-department materials and design information materials are stored in an independent COP (Community of Practice). This established computerized document control system, ANSIM, could lessen a burden for project management team and enhance the productivity as well.

• Journal of Radiation Protection and Research
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• 제46권2호
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• pp.58-65
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• 2021

Background: The evaluation of skyshine distribution, release of airborne radioactive nuclides, and soil activation and groundwater migration were required for radiological assessment of the impact on the environment surrounding In-Flight (IF)-system facility of the RAON (Rare isotope Accelerator complex for ON-line experiment) accelerator complex. Materials and Methods: Monte Carlo simulation by MCNPX code was used for evaluation of skyshine and activation analysis for air and soil. The concentration model was applied in the estimation of the groundwater migration of radionuclides in soil. Results and Discussion: The skyshine dose rates at 1 km from the facility were evaluated as 1.62 × 10^-3 μSv·hr^-1. The annual releases of ³H and ¹⁴C were calculated as 9.62 × 10^-5 mg and 1.19 × 10^-1 mg, respectively. The concentrations of ³H and ²²Na in drinking water were estimated as 1.22 × 10^-1 Bq·cm^-3 and 8.25 × 10^-3 Bq·cm^-3, respectively. Conclusion: Radiological assessment of environmental impact on the IF-facility of RAON was performed through evaluation of skyshine dose distribution, evaluation of annual emission of long-lived radionuclides in the air and estimation of soil activation and groundwater migration of radionuclides. As a result, much lower exposure than the limit value for the public, 1 mSv·yr^-1, is expected during operation of the IF-facility.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.608-617
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• 1999

The powder mixture which has a bimodal size distribution, with a large mode corresponding to AUC-UO$_2$ powder and a small one corresponding to ADU-UO$_2$ powder, was prepared, pressed into compacts, and sintered at 1680t for 4 hours in hydrogen gas. The compact density of the powder mixture increases with increasing ADU-UO$_2$content within a content of 20 wt %, since small ADU-UO$_2$ particles can fill interstices between large AUC-UO$_2$particles. The UO$_2$ pellet made using the powder mixture has a lower open porosity than that made using AUC-UO$_2$ powder alone. The mechanism for the formation of a flake-like pore is proposed, and the decrease in open porosity may be ascribed to the decrease in the number of flake-like pores.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.119-128
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• 2006

We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.662-667
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• 1995

An internal gelation process was adopted for the fabrication of carbon-dispersed UO$_3$ microspheres which will be fed to the fabrication for uranium nitride microsphere fuels by the carbothermic reduction. For investigating the proper process conditions, a composition range of feed solution for preparing good UO$_3$ gel spheres was firstly defined by observing the gelation behavior. Within the defined solution compositions, carbon-dispersed microspheres were prepared and carbon distribution in microspheres were observed by SEM. The results showed that production of good carbon-dispersed microspheres was possible, and the most of carbon were evenly distributed in the microspheres although large carbon-rich aggregates were sparsely existent.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.416-423
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• 1998

Many studies for the perforated plates have been done, especially on the subject of static behavior and stress distribution in the plate. Equivalent elastic properties are one of the successive concepts for this problem. However little effort was taken to get their dynamic characteristics. In this paper finite element modal analysis was performed for the perforated plates having square and triangular hole patterns. An attempt to use existing equivalent elastic properties into the modal analysis of the plate was carried out. To verify feasibility of the finite element models, modal test was also performed on one typical perforated plate. System parameters such as natural frequencies and mode shapes were extracted and compared with the analysis results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.19-24
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• 1997

In this study, we extend the application of the interface-matrix(IM) method for reflector modeling to Analytic Flux Expansion Nodal (AFEN) method. This include the modifications of the surface-averaged net current continuity and the net leakage balance conditions for IM method in accordance with AFEN fomular. AFEN-interface matrix (AFEN-IM) method has been tested against ZION-1 benchmark problem. The numerical result AFEN-IM method shows 1.24% of maximum error and 0.42% of root-mean square error in assembly power distribution, and 0.006%Δk of neutron multiplication factor. This result proves that the interface-matrix method for reflector modeling can be useful in AFEN method.

• Nuclear Engineering and Technology
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• 제33권4호
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• pp.365-374
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• 2001

In order to examine the in-reactor performance of very-high-density dispersion fuels for high flux performance research reactors, U-l0wt.%Mo microplates containing centrifugally atomized powder were irradiated at low temperature. The U-l0wt.%Mo dispersion fuels show stable in- reactor irradiation behaviors even at high burn-up, similar to U$_3$Si$_2$ dispersion fuels. The atomized U-l0wt.%Mo fuel particles have a fine and a relatively uniform fission gas bubble size distribution. Moreover, only one of third of the area of the atomized fuel cross-sections at 70a1.% burn-up shows fission gas bubble-free zones, This appears to be the result of segregation into high Mo and low Mo.