• Journal of the Computational Structural Engineering Institute of Korea
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• v.33 no.1
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• pp.63-72
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• 2020

The purpose of this study is to investigate the effects of the application of various numerical models and frequency contents of earthquakes on the performances of the reactor containment building (RCB) in a nuclear power plant (NPP) equipped with an advanced power reactor 1400. Two kinds of numerical models are developed to perform time-history analyses: a lumped-mass stick model (LMSM) and a full three-dimensional finite element model (3D FEM). The LMSM is constructed in SAP2000 using conventional beam elements with concentrated masses, whereas the 3D FEM is built in ANSYS using solid elements. Two groups of ground motions considering low- and high-frequency contents are applied in time-history analyses. The low-frequency motions are created by matching their response spectra with the Nuclear Regulatory Commission 1.60 design spectrum, whereas the high-frequency motions are artificially generated with a high-frequency range from 10Hz to 100Hz. Seismic responses are measured in terms of floor response spectra (FRS) at the various elevations of the RCB. The numerical results show that the FRS of the structure under low-frequency motions for two numerical models are highly matched. However, under high-frequency motions, the FRS obtained by the LMSM at a high natural frequency range are significantly different from those of the 3D FEM, and the largest difference is found at the lower elevation of the RCB. By assuming that the 3D FEM approximates responses of the structure accurately, it can be concluded that the LMSM produces a moderate discrepancy at the high-frequency range of the FRS of the RCB.

• Journal of Radiation Protection and Research
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• v.22 no.1
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• pp.35-46
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• 1997

A simple and precise method of $^{14}C$ was developed to analyze $^{14}C$ in the environment samples using a commercially available $^{14}CO_2$ absorbent and a liquid scintillation counter. An air sampler and a combustion system were developed to collect HTO and $^{14}CO_2$ in the air and the biological samples simultaneously. The collection yield of $^{14}CO_2$ by the air sampler was in the range of 73-89% . The yield of the combustion system was 97%. In preparing samples for counting, the optimum ratio of $CO_2$ absorbent to the scintillator for mixing was 1:1. No variation of the specific activity of $^{14}C$ in the counting sample was observed up to 70 days after preparation of the samples. The detection limit for$^{14}C$ was 0.025 Bq/gC, which is the level applicable to the natural level of $^{14}C$. The analytical result of $^{14}C$ obtained by the present method were within ${\pm}6%$ of the relative error from the one by the benzene synthesis. The specific activity of $^{14}C$ in the air collected at Taejon during the period of October 1996 ranged from 0.26 to 0.27 Bq/gC. The specific activity of $^{14}C$ in the air collected at 1km from the Wolsong nuclear power plant a 679 MWe PHWR, was $0.54{\pm}0.03$ Bq/gC. The ranges of specific activities of $^{14}C$ in the pine needles and the vegetations from the areas around the Wolsong nuclear power plant were 0.56-0.67 Bq/gC and 0.23-1.41 Bq/gC, respectively.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.12
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• pp.3748-3754
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• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.290-301
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• 1985

600MW(e) CANDU Primary Heat Transport System (PHTS) is composed of the two “figure-of-eight” loops and is designed to operate with the 4% Reactor Outlet Header (ROH) quality at its rated power. This existence of the two compressible regions and the positive flow-qualitly-void feedbacks are the sources of the PHTS flow instability. To ensure the PHTS stability, ROH-ROH interconnect pipes are installed as passive systems. This paper describes the investigation of the PHTS flow instability at its design full power condition. Also studied are the interconnect effect and the inherent system damping effect on the system stability. The time domain stability analyses are accessed by using the ATHER/MOD-I code which is the improved version of the KAERI developed ATHER code. Under the most adverse system modelling, the “figure-of-eight” symmetric loop shows divergent flow oscillations. Under with the interconnect, the PHTS stability is remarkably enhanced so that the system becomes stable. However, even under the conservative pressurizer modelling, the PHTS shows the more convergent flow oscillations. With the interconnect and the pressurizer modelling, its stability is highly credited. Conclusively, the inherent system damping by pressurizer itself can credit the PHTS stability without the interconnect.

• Korean Chemical Engineering Research
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• v.52 no.3
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• pp.279-288
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• 2014

Nuclear energy is expected to meet the growing energy demand while avoiding CO2 emission. However, the problem of accumulating spent fuel from current nuclear power plants which is mainly composed of uranium oxides should be addressed. One of the most practical solutions is to reduce the spent oxide fuel and recycle it. Next-generation fuel cycles demand innovative features such as a reduction of the environmental load, improved safety, efficient recycling of resources, and feasible economics. Pyroprocessing based on molten salt electrolysis is one of the key technologies for reducing the amount of spent nuclear fuel and destroying toxic waste products, such as the long-life fission products. The oxide reduction process based on the electrochemical reduction in a LiCl-$Li_2O$ electrolyte has been developed for the volume reduction of PWR (Pressurized Water Reactor) spent fuels and for providing metal feeds for the electrorefining process. To speed up the electrochemical reduction process, the influences of the feed form for the cathode and the type of anode shroud on the reduction rate were investigated.

• Applied Microscopy
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• v.35 no.3
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• pp.113-119
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with 3.2% of enrichment had been irradiated up to 35,000 MWd/MTU. The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, destructive method analysis has been used but it makes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• Journal of Radiation Protection and Research
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• v.24 no.3
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• pp.131-142
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• 1999

By using the Korean demographic data and the modified relative risk projection model given in the Committee on the Biological Effect of Ionizing Radiation (BEIR) report-V under the U.S. National Academy of Science, the radiogenic excess risk in Korean population has been evaluated. On the basis of this risk, a safety goal for the safe operation of domestic nuclear power plants has been further derived in terms of personal dose. The baseline risk of death due to all causes in Korea and the trivial risk level, which the society considers safe, were estimated to be $5.2{\times}10^{-3}$ and $5.2{\times}10^{-6}$, respectively. The radiogenic excess cancer risk in Korea has been estimated to be $5.2{\times}10^{-3}$ for tie case of acute exposure to 0.1 Gy and $3.7{\times}10^{-3}$ for the case of chronic lifetime exposure to 1.0 mGy/y. On the basis of these risks estimate, the resulting safety goal for one year opeation of a reactor was 0.05 mSv, which is quite identical with the ALARA guideline prescribed by the USNRC in the Appendix I, 10CFR50.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.26 no.2
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• pp.131-138
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• 2013

The high damping rubber bearings contributed to reduce a seismic force transmitted to upper structures, the material properties of rubber changes with time and the rubber with age-related degradation can affect the seismic response of structures and equipments. Therefore the seismic response of structure considering age-related degradation of isolators should be evaluated. In this paper, the stiffness and damping for isolators were defined using the aging data proposed by other researchers. The reactor containment building and the auxiliary building were selected to conduct the nonlinear analysis and the natural frequency, maximum responses, floor response spectrum(FRS) were evaluated with time using the four earthquakes with different frequency contents. According to the analysis results, the seismic responses are increased by the age-related degradation of isolators and the detail inspections should be conducted up to 20 years because it was presented that the change of FRS was high during this period.