• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1318-1326
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• 2020

Structural modification in the electrical cabinet is investigated by a proposed procedure that comprises of an experimental, analytical and numerical solution. This research emphasizes the linear dynamic analysis of the cabinet that is studied under the seismic excitation to demonstrate the real behavior of the cabinets in NPP. To this end, an actual electric cabinet is experimentally tested using an impact hammer test which reveals the fundamental parameters of the cabinet. The Frequency-domain decomposition (FDD) method is used to extract the dynamic properties of the cabinet from the experiment which is then used for numerical modeling. To validate the dynamic properties of the cabinet an analytical solution is suggested. The calibrated model is analyzed under the floor response obtained from the Connecticut nuclear power plant structure excited by Tabas 1978 (Mw 7.4) earthquake. Eventually, the grouping effect of the cabinets is proposed which represents the influence on the dynamic modification. This grouping of the cabinets is described more sophisticatedly by the theoretical understating, which results in a significant change in the seismic response. Considering the grouping effects will be helpful in the assessment of the real seismic behavior, design, and performance of cabinets.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.1-11
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• 2001

In contrast with conventional fieldbus researches which are focused merely on real-time performance, this study aims to evaluate the real-time performance of the communication system including fault-tolerant mechanisms Maintaining performance in presence of recoverable faults is very important in case that the communication network is applied to a highly reliable system such as next generation Nuclear. Power. Plant (NPP). If the tie characteristics meet the requirements of the system, the faults will be recovered by fieldbus recovery mechanisms and the system will be safe. If the time characteristics can not meet the requirements, the faults in the fieldbus can propagate to the system failure. In this study, for the purpose of investigating the time characteristics of fieldbus, the recoverable faults are classified and then the formulas that represent delays including recovery mechanisms are developed. In order to validate the proposed approach, we have developed a simulation model that represents the Korea Next Generation Reactor (KNGR) NSSS Process Control System (NPCS). The results of the simulation show us the reasonable delay characteristics of the fault cases with recovery mechanisms. Using the simulation results and the system requirements, we also can calculate the failure propagation probability from fieldbus to outer system.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3849-3854
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• 2022

The present work aims to optimize the radiation protection efficiency for ion-selective containers used in the liquid treatment for the nuclear power plant (NPP) cooling cycle. Some naturally occurring rocks were examined as filler materials to reduce absorbed dose and equivalent dos received from the radioactive waste container. Thus, the absorbed dose and equivalent dose were simulated at a distance of 1 m from the surface of the radioactive waste container using the Monte Carlo simulation. Both absorbed dose and equivalent dose rate are reduced by raising the filler thickness. The total absorbed dose is reduced from 7.66E-20 to 1.03E-20 Gy, and the equivalent dose is rate reduced from 183.81 to 24.63 µSv/h, raising the filler thickness between 0 and 17 cm, respectively. Also, the filler type significantly affects the equivalent dose rate, where the redorded equivalent dose rates are 24.63, 24.08, 27.63, 33.80, and 36.08 µSv/h for natural rocks basalt-1, basalt-2, basalt-sill, limestone, and rhyolite, respectively. The mentioned results show that the natural rocks, especially a thicker thickness (i.e., 17 cm thickness) of natural rocks basalt-1 and basalt-2, significantly reduce the gamma emissions from the radioactive wastes inside the modified container. Moreover, using an outer cementation concrete wall of 15 cm causes an additional decrease in the equivalent dose rate received from the container where the equivalent dose rate dropped to 6.63 µSv/h.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4679-4683
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• 2022

Environmental radiation monitoring is required to protect from the effects of radiation in industrial fields such as nuclear power plant (NPP) monitoring, and various gamma camera systems are being developed. The purpose of this study is to optimize parameters of a diverging collimator composed of pure tungsten for compactness and lightness through Monte Carlo simulation. We conducted the performance evaluation based on spatial resolution and signal-to-noise ratio for point source and obtained gamma images and profiles. As a result, optimization was determined at a collimator height of 60.0 mm, a hole size of 1.5 mm, and a septal thickness of 1.0 mm. Also, the full-width-at-half-maximum was 3.5 mm and the signal-to-noise ratio was 53.5. This study proposes a compact 45° diverging collimator structure that can quickly and accurately identify the location of the source for radiation monitoring.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.15-28
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• 2019

In accordance with the decommissioning plan for the Kori Unit 1 NPP, the reactor coolant system will be chemically decontaminated as soon as possible after permanent shutdown. This study developed the chemical decontamination process though the development project of decontamination technology of reactor coolant system and dismantled equipment for NPP decommissioning, which has been carried out since 2014. In this study, Oxidation/reduction process was conducted using system decontamination process development equipment of lab scale and was divided into unit and continuous processes. The optimal process time was derived from the unit process, and decontamination agent and the number of process were derived through the continuous processes. Through the unit process, the oxidation process took 5 hours and the reduction process took 4 hours. As optimum decontamination agent, the oxidizing agent was $200mg{\cdot}L^{-1}$ Permanganic acid + $200mg{\cdot}L^{-1}$ Nitric acid and the reducing agent was $2000mg{\cdot}L^{-1}$ Oxalic acid. In the case of the number of processes, all oxide films were removed during the two-cycle chemical decontamination process of STS304 and SA508. In the case of Alloy600, all oxide films were removed when chemical decontamination was performed for three cycles or more.

• Journal of computational fluids engineering
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• v.10 no.3 s.30
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Journal of the Korea institute for structural maintenance and inspection
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• v.28 no.4
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• pp.48-54
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• 2024

NPP (Nuclear power plant) structures have been constructed near to the sea shore line for cooling water and exposed to steel corrosion due to chloride attack. Regarding NPP structures built in the UAE, chloride transport may be more rapid than those in the other regions since the temperature near to the coast is high. In this study, concrete samples with 5,000psi (35MPa) design strength grade were manufactured with the materials and mix proportions, which were the same as used in the UAE NPP structures, then chloride diffusion coefficients were evaluated considering temperature and curing age. The compressive strength and the diffusion coefficient were evaluated and analyzed for the samples with 28 and 91 curing days. In addition, chloride diffusion tests for 91-day-cured condition were carried out in the range of 20℃ to 50℃. The activation energy was obtained through converting the temperature slope to a logarithmic function and it was compared with the previous studies. The proposed activation energy can be useful for a reasonable durability design by using actual temperature-dependent chloride diffusion coefficient.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.04a
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• pp.209-216
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• 2004

Numerical analysis is carried out to identify the appropriateness of the design codes that is available for the tensile design of fastening system at Nuclear Power Plant (NPP) in this study. This study is intended for the cast-in-place anchor that is widely used for the fastening of equipment in Korean NPPs. The microplane model and the elastic-perfectly plastic model are employed for the quasi-brittle material like concrete and for the ductile material like anchor bolt as constitutive model for numerical analysis and smeared crack model is employed for the crack and damage phenomena. The developed numerical model is verified on a basis of the various test data of cast-in-place anchor. The appropriateness of both ACI 349 Code and CCD approach of CEB-FIP Code is evaluated for the tensile design of cast-in-place anchor and it is proved that both design codes give a conservative results compared with real tensile capacity of cast-in-place anchor.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.464-470
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• 2005

The numerical analysis is carried out to identify the influence of design factors to shear capacity of cast-in-place (CIP) anchor in ACI 349 Code that is available for the design of fastening system at Nuclear Power Plant (NPP) in this study. The MASA program is used to develop the numerical analysis model and the developed numerical analysis model is verified on a basis of the various test data of CIP anchor. Both $l/d_o$ and $c_1/l$ we considered as design factors. As a result, the variation of $l/d_o$ has no influence on the shear capacity of CIP anchor but $c_1/l$ has a large influence on the shear capacity of CIP anchor, Therefore, it is proved that ACI 349 Code may give a non-conservative results compared with real shear capacity of CIP anchor according to $c_1/l$.